November 28, 2013

Scattering of diffracting beams of electron cyclotron wavesby random density fluctuations in inhomogeneous plasmas

Hannes Weber, Omar Maj and Emanuele Poli

Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Boltzmannstrasse 2,D-85748 Garching b. Munchen, Germany.

e-mail: omaj@ipp.mpg.de

The propagation of high frequency waves in a random medium is a difficultproblem with applications in a broad variety of fields; the reviews by Ryzhik,Papanicolaou and Keller [1] and by Bal [2] provide a comprehensive and generaloverview of the derivation kinetic radiative transfer models in random mediabased upon the Wigner transform.

In fusion plasma physics in particular, the scattering of beams by turbulentdensity fluctuations has been addressed since the ’80s for lower hybrid waves[3], but it was not considered a concern for electron cyclotron waves until re-cently when calculations for large machines such as ITER became necessary.Tsironis et al. [4] have estimated the effect of random density fluctuations onthe power deposition profile for electron cyclotron current drive beams in ITERby means of a simple Fokker-Planck model; their findings alerted the commu-nity that for large machines such effects might be sizable and triggered furtherinvestigations for heating and current drive [5, 6, 7] as well as diagnostics [8].Nonetheless, a complete modeling including plasma inhomogeneity, short-scaledensity fluctuations, diffraction effects, and dispersive absorption is up to nownot available.

In this work, the Wigner transform and Weyl symbol calculus have beenapplied to the relevant integro-differential equation for the wave electric fieldon the line of the work of McDonald [9], thus obtaining a kinetic radiativetransfer model which is general enough to account for the desired effects. Anappropriate numerical scheme has been put forward and implemented in thenew code WKBeam [10]. Our numerical results show a significant broadeningof the power deposition profile in ITER, while scattering effects are found to benegligible in medium-size tokamaks such as ASDEX upgrade.

[1] L. Ryzhik, G. Papanicolaou and J. B. Keller, Wave Motion 24, 327–370 (1996).

[2] G. Bal, Wave Motion 43, 132–157 (2005).

[3] P. T. Bonoli and E. Ott, Phys. Fluids 25, 359 (1982).

[4] C. Tsironis et al., Phys. Plasmas 16, 112510 (2009).

[5] K. Hizanidis et al., Phys. Plasmas 17, 022505 (2010).

[6] Y. Peysson et al., Plasma Phys. Control. Fusion 53, 124028 (2011).

[7] A. Ram et al., Phys. Plasmas 20, 056110 (2013).

[8] E. V. Sysoeva et al., Plasma Phys. Control. Fusion 55, 115001 (2013).

[9] S. W. McDonald, Phys. Rev. A 43, 4484 (1991).

[10] H. Weber, IPP report 5/134 (2013);http://edoc.mpg.de/display.epl?mode=doc&id=670316&col=33&grp=1311

Influence of density fluctuations on the O–X mode conversion and on

microwave propagation

A. Köhn1, T. Williams2, R. Vann2, E. Holzhauer1, J. Leddy2, M. O’Brien3, M. Ramisch1

1 IGVP, University of Stuttgart, Stuttgart, Germany2 York Plasma Institute, Department of Physics, University of York, York, U.K.

3 EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, U.K.

Electromagnetic waves in the microwave regime are commonly used to both heat plasmas

and diagnose them. If the plasma frequency exceeds the corresponding microwave frequency,

the plasma is referred to as over-dense and becomes inaccessible for this frequency. One way

to overcome this limitation is to use electron Bernstein waves (EBWs): they are of electrostatic

nature and have no high-density cutoff. Additionally, they are very well absorbed at the electron

cyclotron resonance layer and its harmonics. The EBWs can be coupled to injected electromag-

netic waves at the plasma boundary. The efficiency of this coupling process depends on the

normalized density gradient length k0Ln = k0 · ne/|∇ne| (with k0 the vacuum wave number of

the injected microwave and ne the plasma density) and the injection angle with respect to the

background magnetic field. Strong gradients at the plasma boundary in the density and tempera-

ture profiles usually drive turbulence. The resulting perturbations in the density can significantly

distort a traversing microwave beam, reducing the coupling efficiency to the X-mode and thus to

the EBW. Full-wave simulations are performed to investigate and quantify the influence of den-

sity perturbations on this coupling process and on traversing microwaves in general. First, the

deformation of a propagating microwave beam by a single Gaussian-shaped density structure,

reminiscent of a blob or a filament, is presented and discussed as a function of the absolute den-

sity value and of the spatial size of the perturbation [1]. The investigations are then extended

to the influence on the coupling process. Broadband density fluctuations are also considered,

generated with the fluid code BOUT++ [2].

The plasma-microwave interaction is simulated with two FDTD full-wave codes, one treating

a 2D geometry and the other a 3D geometry. Excellent agreement is found for geometries that

are basically two dimensional in nature.

References

[1] T. Williams et al., submitted to Plasma Phys. Control. Fusion

[2] B. D. Dudson et al., Plasma Phys. Control. Fusion 53, 054005 (2011)

Scattering of ECRF waves by edge density fluctuations and blobs

A. K. Ram1 and K. Hizanidis2

1Plasma Science and Fusion Center, Massachusetts Institute of Technology,

Cambridge, Massachusetts 02139, USA 2National Technical University of Athens (part of HELLAS), School of Electrical and Computer Engineering,

Association EURATOM-Hellenic Republic, GR-15773, Greece

Abstract The edge region of tokamak plasmas is replete with spatially distributed density fluctuations and localized turbulent structures such as blobs. Radio frequency (RF) waves, commonly used for heating and for current profile control, have to propagate from the excitation structures to the core of the plasma through this active region. The fluctuations modify the propagation properties of the waves through reflection, refraction, and diffraction. We have been studying the scattering of RF waves by fluctuations and by blobs using a full-wave theory [1]. We assume that edge plasma is cold. The blobs are taken to be either spherical or cylindrical in shape and are treated differently from fluctuations which are considered to be spatially distributed along planar fronts. At the edges of the fluctuations and the blobs we need to satisfy the electromagnetic boundary conditions that follow from Maxwell's equations. These boundary conditions necessarily require the simultaneous excitation of the two, independent, cold plasma waves. Thus, for example, in the electron cyclotron range of frequencies, if an ordinary wave is coupled to the plasma from an external source, the fluctuations and the blobs will not only scatter the ordinary wave but also couple some of the power to the extraordinary wave. The theoretical approach that we follow is similar to that for Mie scattering of electromagnetic waves by dielectric particles. The plasma, both inside and outside the blobs, is assumed to be homogeneous with arbitrary densities in either region; thus, we are not limited to small density fluctuations. The anisotropy induced by the magnetic field is such that the propagation characteristics and the polarization of the RF waves depend on the polar angle with respect to the direction of the magnetic field. The scattering broadens the spectrum of the waves propagating into the core of the plasma. We will present results on the effect of fluctuations and blobs on short wavelength electron cyclotron waves. A comparison with lower frequency, longer wavelength lower hybrid and ion cyclotron waves will be used to illustrate the properties of the scattering process.

Acknowledgment This work is supported by the US Department of Energy, EUROFUSION Consortium, and Association EURFUSION-Hellenic Republic.

[1] A. K. Ram, K. Hizanidis, and Y. Kominis, Phys. Plasmas 20, 056110 (2013).

Closure of the single fluid magnetohydrodynamic equations

in presence of electron cyclotron current drive

E. Westerhof, J. Pratt, and B. Ayten

FOM Institute DIFFER, Dutch Institute for Fundamental Energy Research,

3430 BE Nieuwegein, The Netherlands, www.differ.nl

In a recent paper [1], Hegna and Callen sketch a general procedure for describing RF current

drive effects in single fluid MHD. In addition to the localized EC power deposition in the energy

balance equation, the effect of RF wave absorption appears in Ohm’s law in two terms: in

a quasi-linear, parallel force localized in the region of power deposition and in the electron-

ion friction, i.e. the resistivity. Because EC driven quasi-linear diffusion is dominantly in the

direction of perpendicular momentum, this quasi-linear, parallel force is negligible. The effect

of ECCD is then entirely contained in the closure relation for the resistivity. Assuming a linear

response of the plasma to the different driving forces, we may write J = JΩ +JECCD where JΩ

is the current density from the parallel electric field and JECCD is the EC driven current density.

This implies the modified Ohm’s law as commonly used: E+ v×B = η(J− JECCD) with the

usual (neoclssical) Spitzer resistivity η , which is unmodified by ECCD, while a separate closure

relation then is used to describe JECCD as, for example, in [2].

In this contribution we present a derivation of a new closure relation for JECCD that faithfully

represents the nonlocal character of the driven current and its origin in the Fisch-Boozer mech-

anism [3]. Employing appropriate approximations of the governing kinetic equation, a set of

equations is obtained for ‘current densities’ J1,2 representing the perturbation of the momentum

distribution function at the resonant parallel velocity in the region of perpendicular velocities

below and above the thermal velocity, respectively. The EC driven current then is given by the

balance of J1 and J2. The final model is provided by the following set of equations:

JECCD = J1 +J2, with∂Ji

∂ t=−νiJi− v‖,res∇‖ ·Ji∓SECCD, for i = 1,2 (1)

where ν1,2 represents the collision frequency at the perpendicular velocity associated with J1

and J2, respectively. The sign in front of the driving term SECCD is negative for i= 1 and positive

for i= 2. Results of this model will be compared to solutions of the full Fokker-Planck equation.

References[1] C.C. Hegna, J.D. Callen, Phys. Plasmas 16 112501 (2009)[2] G. Giruzzi et al., Nucl. Fusion 39 107 (1999)[3] N.J. Fisch, A.H. Boozer, Phys. Rev. Lett. 45 720 (1980)

How to do more with ECE noise?

Matthijs van Berkel, Marco de Baar, Dick Hogeweij,

Hugo van den Brand, Hans Zwart, Gerd Vandersteen

January 31, 2014

ECE measurements play an important role in determining indirect quan-

tities such as the sawtooth period, NTMs, thermal diusivity, etc. However,

due to noise and other uncertainties in the ECE-measurements these indirect

quantities become also uncertain. In this presentation, we will show that by

studying and using the stochastic properties of ECE noise we can arrive at

better estimates of these indirectly measured quantities. We will use two

examples, real-time detection of the period of the sawtooth instability and

the estimation of the thermal diusivity based on ECE-measurements.

Session: ECE or Theory

Preference: Oral

1

Computation of the Spitzer function in stellarators and tokamaks with finite

collisionality

W.Kernbichler1, S.V.Kasilov1 2, G.Kapper1, N.B. Marushchenko3

1Association EURATOM-ÖAW, Institute for Theoretical and Computational Physics, TU Graz,

Austria 2Institute of Plasma Physics, National Science Center “Kharkov Institute of Physics and

Technology”, Ukraine 3Max–Planck–Institut für Plasmaphysik, EURATOM Association, Greifswald, Germany

First author email address: winfried.kernbichler@tugraz.at

The generalized Spitzer function is an important part of current drive calculations. In the long mean

free path regime this function can be effectively computed by bounce averaged methods. In regimes

with finite collisionality, the evaluation of this function for stellarator configurations is rather

difficult. Often it is performed using simplified collision models and/or simplified configurations.

In the present report, a parallel version of the kinetic equation solver NEO-2 [1] is used to solve the

Spitzer problem in stellarator geometry. This code does not use any simplifying assumptions for the

collision operator or for the device geometry. Prior to the parallelization of the code, a sequential

NEO-2 version was mainly used for computation of mono-energetic neoclassical transport

coefficients in stellarators [2] and for computation of the generalized Spitzer function in tokamaks

[3], while computations of this function for stellarators required rather long computing times.

In this report, the specific features of the generalized Spitzer function in the case of finite plasma

collisionality are demonstrated for tokamak and stellarator geometry. These features can

significantly modify ECCD in the case of advanced scenarii with reduced wave absorption such as

X3 and O2.

References

[1] Kernbichler W et al 2008 Journal of Plasma and Fusion Research 3 51061

[2] Beidler C D et al 2011 Nuclear Fusion 51 076001

[3] Kernbichler W et al 2010 Contributions to Plasma Physics 50 761

ECRH scenario with selective heating of trapped/passing electrons in the W7-X Stellarator

N. B. Marushchenko, V. Erckmann, C. D. Beidler, J. Geiger, H. P. Laqua,P. Helander, H. Maassberg, Y. Turkin

Max Planck Institute for Plasma Physics, EURATOM Association, Wendelsteinstr. 1, D-17491 Greifswald, Germany

Using specific features of 'high-mirror' magnetic equilibria for the W7-X Stellarator, ECRH scenarios with the possibility of selective heating of either passing or trapped electrons are discussed. Since the numerical tools developed to date for modelling plasma heating and transport physics cover such a scenario only partly, this kind of experiment can be particularly interesting and useful for investigating ECRH/ECCD kinetics as well as transport physics.

The standard scenario in W7-X calls for injection from the main ports of up to ten RF beams (1 MW each at 140 GHz) for ECR heating and current drive with X2-mode at 2.5 T. The RF power can be launched from the main ports, located near the “bean-shaped” plane (A- and E-ports), where the magnetic field strength B has a maximum and a strong tokamak-like magnetic field gradient. Additionally, two beams can be switched to the ports near the “triangular plane” (N-ports), where B has a minimum and the magnetic field gradienti s weak. For two of the N-ports of W7-X, "remote-steering" launchers are foreseen [1].

For the 'high-mirror' magnetic configuration with Bmax/Bmin = 1.75 and the fraction of trapped particles about 0.6, the cyclotron absorption of the X-mode can appear at the 2nd harmonic for the beams near the maximum of B and at the 3rd harmonic for the beams near the minimum of B. Most important, however, is that the power is deposited into the different classes of electrons, X2 absorption by passing and X3 by trapped electrons.

In the present report, we consider a scenario with pre-heating of the plasma by the X2 mode launched from the A- and E-ports (five beams with total power 5 MW) and then replacing two beams by those launched from the N-ports. For a density 5×1019 m-3 and the temperature of the preheated plasma about 5 keV, predicted by transport simulations, the single-path absorption of the X3-mode is expected to be close to 100%. Usually, the deposition profile for X2 is much narrower than that for X3, but for very oblique launching of X2 beams it is possible to make the deposition profiles for them very similar to the X3-beams launched from the N-ports. The main difference in this case is the different populations of the electrons responsible for absorption. As a consequence, the rapid switch of power from the AE-ports to the N-ports produces a significant changes of the ECRH/ECCD kinetics with practically the same deposition profiles.

References1. W. Kasparek et al., this conference.

On the criteria guiding the design of the upperelectron-cyclotron launcher for ITER

E. Poli1, C. Angioni1, F. J. Casson1, D. Farina2, L. Figini2, T. Goodman3, O. Maj1,O. Sauter3, H. Weber1, H. Zohm1, G. Saibene4, M. Henderson5

1Max-Planck-Institut fur Plasmaphysik, Garching, Germany2Istituto di Fisica del Plasma CNR, EURATOM-ENEA-CNR Association, Milano, Italy3Centre de Recherches en Physique des Plasmas, CRPP-EPFL, Lausanne, Switzerland

4Fusion for Energy, Barcelona, Spain5ITER Organization, Saint-Paul-Lez-Durance, France

The main goal of the Upper Launcher for electron cyclotron waves in ITER is thestabilization of MHD instabilities, in particular Neoclassical Tearing Modes (NTMs) [1].Simple criteria for NTM suppression have guided the choice of the beam parameters.In particular, the request that the driven current jCD should exceed by a factor 1.2 theunperturbed bootstrap current on the relevant rational surface [2] has lead to the design ofvery well focused beams. It has been pointed out, however, that reducing the width of thejCD-profile below the marginal magnetic-island width is not further beneficial for NTMstabilization, so that the total driven current should be rather maximized assuming thatthe profile width remains below a threshold of about 5 cm for ITER [3]. For this reason, anincrease of the envisaged toroidal angle has been proposed for the lower steering mirror ofthe launcher [4]. On the other hand, there are several effects which can potentially broadenthe the jCD profiles calculated by beam tracing codes, like aberration, incomplete beamsuperposition, radial transport of fast (current carrying) electrons, beam scattering due todensity fluctuations, etc. We re-derive the stabilization criteria from a simple form of theRutherford equation to assess their limit of validity, and analyze the relative importanceof different beam-broadening effects. In particular, we investigate the diffusion of fastelectrons by means of gyrokinetic simulations in the absence of the island, to find anupper limit to the expected transport level. Recent results on beam propagation in thepresence of density fluctuations [5] suggest beam scattering to be probably the mostdeleterious cause of profile broadening in big machines. The role of beam misalignmentwith respect to the target surface and the impact of broader profiles on NTM stabilizationin ITER are discussed.

References1) D. Strauss et al., Fus. Eng. Des. 88 (2013), 2761.2) H. Zohm et al., Plasma Phys. Control. Fusion 49 (2007), B341.3) O. Sauter et al., Plasma Phys. Control. Fusion. 52 (2010), 025002.4) N. Bertelli et al., Nuclear Fusion 51 (2011), 103007.5) O. Maj et al., this conference.

Recent physics studies using ECH/ECCD at the 2nd and 3rd harmonics in TCV

T.P. Goodman1, for the TCV team

1 Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), CH-1015

Lausanne, Switzerland

The Tokamak à Configuration Variable, TCV, has used Electron Cyclotron Heating and Current Drive as its only auxiliary heating system for nearly two decades. EC has the invaluable ability to adapt to the extreme range of plasma shapes and positions used within the TCV vacuum vessel, thanks to 7 independently controlled launching antennas and 9 mm-wave sources (6 at 82.7GHz and 3 at 118GHz, each at 0.5MW and 2s pulse duration). TCV’s modest magnetic field (≤1.43T) results in a cut-off density for X2 heating of 4•1019m-3; however, X3 top-launch absorption remains available and can be used effectively at nearly twice that density. X2 heating near the edge is still possible for some high-density configurations and has been used to great effect, e.g. for ELM control. Recent experiments ranged from studies of novel snowflake divertor shapes to mitigate wall loading in H-mode plasmas heated by X2 & X3, to developing the unique current profiles required to produce 3D structures in the core of plasmas having an axisymmetric boundary, such as those possible in the ITER hybrid scenarios, to investigations of electron internal transport barriers, neoclassical tearing mode prevention, sawtooth control and ELM control. Some of these experiments will be described here. The flexibility of the launching antennas is matched by an identical receiving antenna fitted with a set of fast polarizers and used for oblique ECE, Doppler backscattering, correlation ECE (CECE)1 or double-pass transmission measurements2. The latter complements measurements from newly installed Hard X-Ray cameras (HXRS)3 (able to be oriented to look either in or out of the poloidal plane) to study quasi-linear effects during ECH4. Similarly, Doppler backscattering, tangential phase contract imaging (TPCI) and CECE were combined in the identification and investigation of GAMs in the edge of TCV plasmas1.

Building on previous open and closed-loop control of NTMs and sawteeth using EC launchers, control of plasma instabilities now benefits from real-time knowledge of the equilibrium using the code RT-LIUQE. Robust NTM control, in which a sinusoidal variation of the actuator deposition around the target q-surface is added to the tracking, is shown to be more effective at stabilizing the NTM modes as the plasma evolves in time4. Again building on past experience with successful sawtooth control by either pacing or locking, measurements of the effect of sawteeth on the plasma rotation with high effective time resolution – as a function of the phase in the sawtooth cycle – have been made5 using coherent averaging. The installation of diagnostics in ports directly adjacent to the EC inputs, and the regular use of computer controlled feedback, demand a robust protection against stray EC power. Such a system is now in operation. The fast-ion tails in the ion distribution function that are generated by ECCD have been studied at higher power than in the past with HXRS and NPA - in both oblique and radial viewing orientations - providing stronger support for the interpretation6 that these tails arise from Ion Acoustic Turbulence.

The TCV tokamak is essential in the implementation of the European fusion Roadmap7 towards the realization of fusion energy by 2050 and is presently being upgraded to include a 1MW NBI source and an opening for a second future source, together with two 1MW X3 gyrotrons. The option of designing the gyrotrons as dual frequency tubes providing either 126GHz or 84GHz output (for X2 heating) is being explored. Following the recent (2013) end-of-life loss of several X2 gyrotrons, 0.75MW replacements are planned at the present frequency. The expected evolution of the TCV EC system over the next 4 years will be outlined.

[1] L.Porte et al. this conference. [2] T.P.Goodman et al., 24th IAEA Fusion Energy Conference, San Diego, USA 8-13 October 2012. [3] J. Kamleitner, et al., 40th European Physical Society Conference on Plasma Physics and Controlled Fusion, Espoo, Finland, 1-5 July 2013, p. P5.127. [4] J. Decker et al., this conference. [5] B.P. Duval et al., 21st European Fusion Physics Workshop 2013, 9th – 11th December 2013, Ringsted, Denmark. [6] Christian Schlatter, Turbulent ion heating in TCV tokamak plasmas, Ph.D. thesis, no. 4479, Ecole polytechnique fédérale de Lausanne (EFPL), CH-1015 Lausanne, Switzerland, July 2009. [7] http://www.efda.org/wpcms/wp-content/uploads/2013/01/JG12.356-web.pdf

High power ECRH and ECCD in moderately collisional ASDEX Upgrade H-modes

and status of EC system upgrade

J. Stober, A. Bock, E. Fable, F. Sommer, C. Angioni, F. Leuterer, F. Monaco,

S. Müller, M. Münich, B. Petzold, E. Poli, M. Schubert, H. Schütz, D. Wagner, H. Zohm

and the ASDEX Upgrade Team

Max-Planck-Institut für Plasmaphysik, Garching, Germany

A. Meier, Th. Scherer, D. Strauß, J. Jelonnek, M. Thumm

Karlsruhe Institute of Technology, EURATOM-Association, Karlsruhe, Germany

W. Kasparek, B. Plaum

IGVP, Universität Stuttgart, Stuttgart, Germany

A. Litvak, G.G. Denisov, A.V. Chirkov,

Institute of Applied Physics, RAS, Nizhny Novgorod, Russia

E.M. Tai, L.G. Popov, V.O. Nichiporenko, V.E. Myasnikov, E.A. Soluyanova, V. Malygin

GYCOM Ltd., Nizhny Novgorod, Russia

In 2011 the first upgrade of ECRH systems of ASDEX Upgrade was largely complete (although

for several reasons only 3 of the four new gyrotrons were so far operational at the same time).

Since then almost 4 MW of ECRH power in the plasma were available, delivered from 7 gyrotrons

at 140 GHz. Here we report on experiments replacing NBI heating by ECRH, i.e replacing some

ion heating by electron heating. The early experiments reported already at EC17 used moderate

heating powers of approximately 3 MW in order to be able to supply all the heating power also by

ECRH alone, resulting in central temperatures of only 1-2 keV and correspondingly high

collisional heat exchange between electrons and ions. The heat transfer due to this equilibration

process to some extend counteracted the effects of heating the one or the other species. Still it was

obvious that replacing only 30% of the NBI by ECRH lowered the central Ti significantly.

Consequently the experiment was repeated at higher heating powers (8 MW), higher temperatures,

somewhat reduced density and significantly lower collisionality. Up to 45 % of this power could

be supplied by ECRH and the negative effect of electron heating on Ti became even stronger. The

results were compared with TGLF [1] modelling. The agreement between experiment and TGLF

was generally very good, not only in the trend but also in absolute numbers, using a Kadomtsev-

type sawtooth (ST) model. The results demonstrate the crucial role of Te/Ti for the ITG dominated

transport over a wide range of Te/Ti. A small increase of Te by ECRH decreases the ITG stability

leading to a reduction of Ti and a further increase of Te/Ti such that ion transport is further

increased. This effect became also obvious when doing central ctr-ECCD experiments in order to

achieve q-profiles with flat central shear and qmin>1 at even lower collisionality (lower current).

With 3.5 MW of ECCD and 5MW of NBI central temperatures of Te = 6 keV and Ti = 2 keV were

obtained. Adding 7.5 MW of NBI brought Ti up to Te, almost doubling the H-factor from 0.8 to 1.5.

Also this behavior is quantitatively modeled very well with TGLF (here without ST, qmin > 1). For

larger values of νei ⋅τE, as expected in ITER, the choice of the heating channel will be less crucial.

To enable studies with higher power and pulse length on AUG, the first ECRH system will be

replaced using the same launching positions with slightly modified launchers. The upgrade

roughly doubles the power of these four units and extends their pulse length to 10s, such that

finally around 6.5 MW at 140 GHz (or 5.5 MW at 105 GHz) will be available in the plasma from

8 units during the whole AUG discharge. Status and plans for this upgrade will be reported as well.

[1] Staebler, G.M. et al, Physics of Plamas, 14, 055909 (2007)

APPLCATION OF O- AND X-ECE FOR INVESTIGATION OF INITIAL STAGE OF

DISCHARGE IN T-10 TOKAMAK

V.I. Poznyak, V.V. Pitersky, G.N. Ploskirev, E.G. Ploskirev

NRC “Kurchatov Institute”, IFT, 123182, Kurchtov Sq. 1, Moscow, Russia

Formation of the initial stage of the discharge and current profile management for ITER

are important objectives for the ECH method. However, a role of electron plasma oscillations in

creation of electron distribution function does not take into consideration in models of possible

scenarios. Already in the first stage, certain shape of distribution function installs and remains in

quasi-stationary stage as it was shown earlier [1, 2].

This work continues the analysis of the kinetic characteristics of plasma creating by the

only inductive method and also with using of ECH. Ordinary (1st harmonic, 59-78 GHz) and

extraordinary polarization (2nd

harmonic, 118-164 GHz) are used for the profile ECE

measurements. ECE of high energy electrons is measured in the downshifted band of 1st

harmonic (37-58 GHz). This allows determine the characteristics of the tail part of distribution

function in the central region of column, specified by the potential wave process. Plasma noises

are recorded into band 0.5-16 GHz. The time of the current wave propagation inward plasma is

defined by the appearance of oscillations and instabilities on q=1, 3/2, 2.

The velocity of the current penetration inwards of the plasma column (pinch-wave) is

investigated dependently on the level of impurities, arising at time of breakdown, and value of

electric field, forming by inductor. It was discover that current pinch velocity grows nonlinear

(exponentially) with value of the applied electric field. Similar dependence of pinch-wave is

observed under an enhancement of impurity (carbon) amount, incoming into discharge just in

breakdown. Dynamics of the plasma noise power has the same dependence. Pinch velocity

increases under growth of electron density and decreases under magnetic field rise. Experimental

data show to that maximal fast compression of the current channel occurs in the internal hot and

dense region.

Complex of data shows that, for a description of the current channel formation, should

consider more complex parametric dependence of the plasma electric conductivity than its

neoclassical expression. The dependence of the rate of the pinch-wave on the electron

temperature, density, electrical and magnetic fields give grounds to assume the defining role of

potential plasma oscillations in transport of the longitudinal electron momentum, aimed inside

the plasma column at the formation of current channel.

We think it is substantial, in addition to the technical task for ECH – to reduce the load on

the inductor, to solve the physical problem – to limit the speed of the current wave inside the

column in order to avoid the critical value of the internal longitudinal electric field, leading to

excitation of plasma oscillations and instabilities [3].

This work is supported by the Ministry of education and science (Contract No

16.518.11.7004) and Rosatom (the Contract No Н.4f.45.90.11.1021).

1. V.I. Poznyak, T.V. Gridina, V.V. Pitersky at al., http://dx.doi.org/10.1051/epjconf/

20123203010

2. V.I. Poznyak, V.V. Pitersky, G.N. Ploskirev, E.G. Ploskirev, http://dx.doi.org/10.1051/

epjconf/20123203008

3. Poznyak V.I., Bagdasarov A.A., Piterskii V.V. et al., Proc. of 15th

Int. Conf. on Plasma Phys.

and Control. Nucl. Fusion Research, Seville, Spain, 1994, Nucl. Fusion, 2, 169 (1995).

Experimental characterization of plasma start-up using ECRH in preparation of W7-X operation

M. Preynas, S. Kobayashi, S. Kubo, H.P. Laqua, K. Nagasaki, T. Shimozuma, T. Stange, D. Aßmus, H. Igami, S. Kado, T. Mutoh, M. Otte, Y. Yoshimura.

1Max-Planck-Institut für Plasmaphysik (IPP), EURATOM-Association, D-17491 Greifswald, Germany 2Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011, Japan

3National Institute for Fusion Science, 322-6 Oroshi-Cho, Toki-City, Gifu 509-5292, Japan

E-mail: melanie.preynas@ipp.mpg.de

The upcoming operation of W7-X will be supported by an ECRH system working at 140 GHz in 2nd harmonic. Beyond the standard start-up scenario of on-axis heating in X2, off-axis heating and 3rd harmonic are also considered for plasma operation at lower magnetic field to get a further operation point for W7-X. Because the optimization of the plasma breakdown is crucial to ensure a successful plasma build-up, dedicated plasma start-up experiments were performed on three stellarators: Heliotron J, LHD and WEGA. Start-up behavior and dependencies on ECRH injected power, neutral gas pressure and rotational transform have been obtained. In addition, the experiments on Heliotron J have allowed a detailed characterization of the temporal evolution from the breakdown phase to the plasma build-up phase.

Plasma start-up delay time decreases with the increase in ECRH input power. However, this behavior saturates when low pre-fill neutral gas pressure conditions are met. Both the delay time and the electron density once plasma is built-up are an increasing function of the gas pressure. On Heliotron J and WEGA devices, the higher the rotational transform is, the faster the start-up and the higher the plasma density are. Analysis of the temporal evolution of the plasma start-up in on-axis heating and X2 has led to the separation of the plasma start-up in two main phases. The first one is characterized by a thin ring of plasma around the magnetic axis, having a radial size of the order of the ECRH beam size. The snake is sustained until the electron density reaches a value where the ECRH absorption (and hence the electron heating) is significantly efficient to overcome the losses (drift, elastic collisions, recombinations), leading to the increase in electron temperature. First signals in the poloidally distributed lines of Hα measurements attest the occurrence of ionization process throughout the confined region. This provides the evidence of plasma expansion that initiates the plasma build-up phase. Then, both Hα and electron density increase up to a maximum, which defines the end of the whole plasma start-up phase. Additionally, off-axis heating experiments performed in X2 are characterized by a longer plasma start-up duration compared to on-axis heating discharges. Third harmonic in X-mode has been attempted on LHD for different neutral gas puffing settings but no plasma breakdown has been achieved.

This multi-machine study has been useful to define the ECRH start-up scenarios for W7-X. The parametric study has also provided necessary data for the development and the validation of modeling.

Non-inductive Current Start-up by Electron Cyclotron Heating and Current Drive in KSTAR

H. Tanaka a, M. Uchida a, T. Maekawa a, Y-S. Bae b, M. Joung b, J. H. Jeong b

a Graduate School of Energy Science, Kyoto University, Japan

b National Fusion Research Institute, Korea

E-mail address: h-tanaka@energy.kyoto-u.ac.jp

Non-inductive current start-up in tokamaks is required to eliminate or reduce the center solenoid (CS) coils, which is expected to realize compact and economical fusion reactors. Electron cyclotron (EC) heating and current drive (ECH/ECCD) is an attractive candidate among the non-inductive current drive methods because only small launchers behind the plasma facing components are needed. Non-inductive spherical tokamak formation by ECH/ECCD was successfully demonstrated in small devices. However, in large tokamaks with conventional aspect ratio such as DIII-D and JT-60U, the plasma current of only several tens of kA was generated by injecting the microwave power of MW level, and the formation of closed flux surfaces was not clearly achieved. KSTAR is a conventional aspect ratio tokamak (R=1.8m, a=0.5m) and has super-conducting magnets like ITER. When the CS coil currents are kept at zero, the plasma current Ip is started up to 14.5 kA by injecting 84 GHz microwave power of 180 kW for 2 sec and by controlling the equilibrium vertical field strength and its decay index. The microwave power is injected from low field side obliquely to the toroidal field in O-mode. The fundamental EC resonance layer is located at R=1.5m. The soft X-ray signals detected by a 16 ch Si photodiode detector array and the poloidal flux signals from flux loops wound around the inner column near the equatorial plane show significant increase when Ip exceeds ~13 kA, which indicate the formation of closed flux surfaces. The plasma current profile is obtained by least-squares fitting to 39 flux loop signals with a model analytic function. Reconstructed poloidal flux distribution shows that closed flux surfaces are formed at t ~ 1.6 sec when Ip exceeds ~10.5 kA. The size of the last closed flux surface become ~0.4 m at the last stage of the discharge. The cross-field-passing- electron (CFPE) model for formation of closed flux surfaces is applied to the experimental conditions and the result shows that CFPE current of 2.26 kA is driven by 115 kW EC power and produces the closed flux surfaces with addition to the pressure driven current of 10.63 kA. However, the CFPE current does not flow in the closed flux surfaces because the asymmetric confinement in the velocity space disappears there. Then the ramp-up of plasma current after the formation of closed flux surfaces should be caused by EC-driven current in the closed flux surfaces. At the last stage of the discharge, the most probable current profile is obtained by assuming that it is a sum of model current profiles of EC-driven current and pressure-driven current, and by looking for the profile which reproduces the observed poloidal flux well and has the reasonable equilibrium pressure profile. It is found that the EC-driven current of 4.5 kA flows in the closed flux surfaces and the pressure-driven current of 10 kA flows in the open field region at the last stage.

Noninductive formation of an extremely overdense spherical tokamak by electron Bernstein wave heating and current drive on LATE

M. Uchida, T. Maekawa, H. Tanaka, Y. Noguchi

Graduate School of Energy Science, Kyoto University, Kyoto, Japan

Noninductive startup of Tokamak without the use of the central solenoid is a key issue for an economical reactor. In the Low Aspect ratio Torus Experiment (LATE) device, plasma current startup and formation of spherical tokamak by electron Bernstein (EB) waves at an extremely overdense regime has been explored.

The microwaves at 2.45GHz from four magnetrons are injected from midplane four launchers with an oblique angle in the form of O-mode. When a microwave power of Pinj ~ 10 kW is injected under a weak Bv field, a plasma current is initiated and increases, resulting in the formation of closed flux surfaces. Next, the plasma current ramps up with ramps of the microwave power and Bv, and finally reaches Ip = 10.5 kA by Pinj = 58 kW. Equilibrium analyses and hard X-ray measurements indicate that the plasma current is carried by an electron-cyclotron driven fast electron tail with an energy range up to ~100 keV.

The electron density increases as Ip increases and the final line averaged density reaches 7 times the plasma cutoff density. Such an increment of the density is found to be obtained when the ECR layer lies on the plasma core and the upper hybrid resonance layer lies to slightly higher field side of the 2nd ECR layer. In this case, the incident electromagnetic waves are mode converted to EB waves at their first propagation band (between the fundamental and the 2nd ECR layer), after which the EB waves propagate towards the fundamental ECR layer and may heat the bulk electrons as well as the current carrying fast electrons at the plasma core.

The time evolution of extreme ultra violet (XUV) emission signal profile shows a significant increase near the ECR layer as Ip increases, indicating the heating of bulk electrons at the ECR layer. The increase is much larger than the increment in electron density. In addition, impurity line radiations at higher excitation energies such as CV (304 eV) and OV (72 eV) appear and strongly increase compared with the density increment. These suggest that the bulk electron temperature also increases by EB heating.

Long Pulse EBW Start-up Experiments in MAST

V.F. Shevchenkoa, T.S. Bigelow

b, J.B. Caughman

b, S. Diem

b, C. Dukes

b, P. Finburg

a, J.

Hawesa, C. Gurl

a, J.Griffiths

a, J. Mailloux

a, M. Peng

b, A.N. Saveliev

c, Y. Takase

d

a EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

b Oak Ridge National Laboratory, Oak Ridge, TN, USA 37830

c Ioffe Institute, Politekhnicheskaya 26, 194021 St. Petersburg, Russia d University of Tokyo, Kashiwa 277-8561, Japan

Non-inductive plasma current start-up is a critical issue in spherical tokamak (ST)

research because of lack of space for a shielded central solenoid. Various techniques have

been proposed and developed in order to avoid a central solenoid in future ST devices. We

report on the electron Bernstein wave (EBW) start-up technique based on a 28 GHz gyrotron.

The EBW start-up method deployed here relies on the production of low-density plasma by

RF pre-ionization around the fundamental electron cyclotron resonance (ECR). Then a

double mode conversion (MC) is considered for EBW excitation. The scheme consists of MC

of the ordinary (O) mode, incident from the low field side of the tokamak, into the

extraordinary (X) mode with the help of a grooved mirror-polarizer incorporated in a graphite

tile on the central rod. The X mode reflected from the polarizer propagates back to the plasma,

passes through ECR and experiences a subsequent X to EBW MC near the upper hybrid

resonance (UHR). Finally the excited EBW mode is totally absorbed before it reaches the

ECR, due to the Doppler shifted ECR. The absorption of EBW remains high even in cold

rarefied plasmas. Furthermore, EBW can generate significant plasma current during the start-

up phase giving the prospect of a fully solenoid-free plasma start-up.

First experiments using this scheme were carried out on the Mega Amp Spherical

Tokamak (MAST) at Culham Science Centre, UK [1]. Plasma currents up to 33 kA have

been achieved using 100kW (as measured at the gyrotron) 90ms RF pulses without the use of

solenoid flux. It was shown that the plasma formation and current generation are governed

predominantly by EBW current drive. Recently experimental results were extended to longer

RF pulses (up to 0.5 s) showing further increase of plasma currents generated by RF power

alone. A record current of 73kA has been achieved with 70kW (as measured just before the

MAST vessel) 0.45s RF pulse. Experiments with a vertical modulation of the plasma position

with respect to the midplane were conducted to clarify the physical mechanisms responsible

for the current drive. Actual RF power injected into the MAST vessel was about same in both

sets of experiments. The current drive enhancement was mainly achieved by the RF pulse

extension and further optimisation of the start-up scenario.

[1] V. Shevchenko, et al, Nucl. Fusion 50 (2010), 022004.

This work was funded by the RCUK Energy Programme under grant EP/I501045 and the

European Communities under the contract of Association between EURATOM and CCFE.

The views and opinions expressed herein do not necessarily reflect those of the European

Commission.

42GHz ECRH assisted Plasma Breakdown in tokamak SST-1

B. K. Shukla, S. Pradhan, Paresh Patel, Rajan Babu, Jatin Patel, Harshida Patel, Pragnesh Dhorajia, Prashant Singh, R. Jha and D. Bora

Institute for Plasma Research, Bhat, Gandhinagar-382428 Email: shukla@ipr.res.in

Abstract

The Electron Cyclotron Resonance Heating (ECRH) is an important heating system in tokamak to carry out various experiments related breakdown, heating and current drive. In SST-1, 42 ECRH system has been commissioned to carry out breakdown and heating experiments at 0.75T and 1.5T operating toroidal magnetic field. The 42GHz ECRH system consists of high power microwave source Gyrotron, approximately 20 meter long transmission line and a mirror based launcher. The Gyrotron delivers 500kW power for 500ms duration at 50kV beam voltage and 20A beam current. This is a depressed collector type Gyrotron and the efficiency is better than 50%. The Gyrotron consists of internal mode converter and delivers power in TEM00 mode (Gaussian output) with mode purity better than 95%. Initially, the Gyrotron is tested on dummy load, after successful testing at full parameters on dummy load, the system is connected to tokamak using corrugated waveguide based transmission line system. The transmission line consists of matching optic unit, DC break, 63.5mm diameter corrugated waveguides, bend with bi-directional couplers, polarizer and bellow etc. The total transmission loss in the line is less than 1.1dB. The transmission line terminates with a vacuum barrier window, which is connected to tokamak with the help of a UHV gate valve. The ECRH launcher consists of two mirrors one focusing and one plane to launch focused ECRH beam in plasma. The 42GHz system has been commissioned on tokamak SST-1 and ECRH power in fundamental O-mode & second harmonic X-mode is launched from low field side (radial port) of tokamak. At 0.75T operation, approximately 300kW ECH power is launched in second harmonic X-mode and successful ECRH assisted breakdown is achieved at low loop voltage ~ 3V. The ECRH power is launched around 30ms prior to loop voltage. The hydrogen pressure in tokamak is maintained ~ 1x10-5mbar and the pre-ionized density is ~ 4x1012/cc. At 1.5T operating toroidal magnetic field, the ECH power is launched in fundamental O-mode. The ECH power at fundamental harmonic is varied from 100kW to 250kW and successful breakdown is achieved in all ECRH shots. In fundamental harmonic there is no delay in breakdown while at second harmonic ~ 30ms delay is observed, which is normal at second harmonic breakdown. The paper discusses about 42GHz ECRH system on SST-1 and explains the results of ECRH assisted breakdown in tokamak at fundamental and second harmonic.

Application of Electron Cyclotron Heating to the Study of Transport in ITER Baseline Scenario-like Discharges in DIII-D

R.I. Pinsker1, M.E. Austin2, D.R. Ernst3, A.M. Garofalo1, B.A. Grierson4, J.C. Hosea4,

T.C. Luce1, A. Marinoni3, G.R. McKee5, C.C. Petty1, M. Porkolab3, L. Schmitz6, W.M. Solomon4, G. Taylor4, and F. Turco7

1General Atomics, P.O. Box 85608, San Diego, California 92186-5608, USA 2University of Texas at Austin, 1 University Station, Austin, Texas 78712, USA

3Massachusetts Institute of Technology, Cambridge, Massachusetts 02139, USA 4Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543-0451, USA

5University of Wisconsin-Madison, Madison, Wisconsin 53706, USA 6University of California Los Angeles, P.O. Box 957099, Los Angeles, CA 90095, USA

7Columbia University, New York, New York 10027, USA

Recent DIII-D experiments in the ITER Baseline Scenario (IBS) have shown strong increases in fluctuations and correlated reduction of confinement associated with entering the electron-heating-dominated regime with strong ECH. Properties of discharges in the baseline operating scenario for ITER [ITER-similar shape, βN~2, q95~3, ELMing H-mode with H98(y,2)~1] are the subject of ongoing experiments on DIII-D; the recent work has extended the similarities to include ITER-relevant (low) torque and Te/Ti~1 (dominant electron heating). The advantage of electron cyclotron heating as a principal heating tool for these studies is the fine control of the power deposition profile that is possible and the fact that all of the power goes to the electrons allows detailed study of the transport properties of the discharge. The addition of 3.2 MW of 110 GHz EC power at ρ~0.42 to IBS discharges with ~3 MW of neutral beam injection causes large increases in low-k and medium-k turbulent density fluctuations observed with Doppler backscatter (DBS), beam emission spectroscopy (BES) and phase-contrast imaging diagnostics, correlated with decreases in the energy, particle, and momentum confinement times. In cases in which the neutral beam power is feedback controlled to maintain a constant stored energy, study of the dynamics upon turn-off of the ECH power shows reduced confinement during ECH, including the well-known reduction in particle confinement (both deuterium and impurities) often associated with ECH. Power balance calculations show the electron heat diffusivity χe increases significantly in the mid-radius region 0.4<ρ<0.8, which is roughly the same region where the DBS and BES diagnostics show the increases in turbulent density fluctuations. Confinement of angular momentum is also reduced during ECH. Significant differences between the character and frequency of the edge localized modes (ELMs) with and without EC are observed; the more frequent small ELMs obtained during ECH have proven to be a useful application in the ITER baseline scenario studies. An initial linear gyrokinetic analysis of these discharges to identify unstable modes and their growth rates is planned. Physics understanding of the mechanisms behind the confinement changes observed during torque-free, non-fueling pure electron heating may ultimately yield tools to improve control of D-T, alpha-heated, low rotation ITER discharges.

This work was supported in part by the US Department of Energy under DE-FC02-04ER54698, DE-FG03-97ER54415, DE-AC02-09CH11466, DE-FG02-89ER53296, DE-FG02-08ER54999, DE-FG08ER54981, and DE-FG02-04ER54761.

Performance History and Upgrades for the DIII-D Gyrotron Complex

J. Lohr1, M. Cengher1, Y.A. Gorelov1, S. Noraky1, D. Ponce1, A. Torrezan1, E. Kolemen2, R.A. Ellis2, and L. Myrabo3

1General Atomics, P.O. Box 85608, San Diego, California 92186-5608, USA 2Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, NJ 08543-0451, USA

3Lightcraft Technologies, Bennington, VT, USA The gyrotron system on the DIII-D tokamak presently comprises 6 operating tubes, with infrastructure for 8, all of which nominally generate 1 MW for short pulses and ≈800 kW for operational pulse lengths, which are limited administratively to 5 s, consistent with the DIII-D pulse length. All of the gyrotrons operate at 110 GHz. Two are a depressed collector design, part of a recent upgrade, while the others are older undepressed diode tubes. As funding permits, the installation is transitioning to a new series of gyrotrons with a higher frequency of 117.5 GHz, the first example of which has demonstrated 1.8 MW 5 ms pulse operation at 60 A and 1.5 MW at 50 A in initial testing. Two gyrotrons are undergoing repair and will be returned to service. One developed an internal water leak and a second failed due to a problem with the electron gun. Both of these failed tubes had been in service for about 10 years. Maximum injected rf power has been about 3.4 MW and this performance level is expected to continue to be realized during the 2014 experimental campaign.

The gyrotron system on DIII-D began operation with 1 MW class tubes in 1997 and since then has been steadily increased in capability. The overall reliability for successfully meeting a performance request has been about 85% even though four gyrotrons are operated in pairs on single power supplies, hence a fault in one causes two to be lost for the shot.

There are four dual launcher assemblies on DIII-D, each of which can handle full power from two of the gyrotrons, with individual waveguides and fully articulating steering mirror assemblies. Using these steering mirrors, the rf beams can be directed over a 40º range in the vertical, poloidal, plane and ±20º horizontally, giving access to the plasma center and above, while providing current drive in either direction. This steering capability is being upgraded with higher speed drive motors and new position readback encoders, which have demonstrated a full 40º poloidal scan in <200 ms and about 2 mm position accuracy for the rf beam at the plasma center in initial tests. The aiming and appropriate polarizer angles can be set up and fixed prior to a plasma shot with preprogrammed time dependence of the rf power, but aiming and output control can also be assumed by the DIII-D Plasma Control System (PCS). Under PCS control, real time equilibrium calculations guide the aiming. Feedback control, based on diagnostic input, provides the basis for determining the required heating or current drive and the result is the ability to fully suppress MHD instabilities such as the neoclassical tearing mode, to influence the j(r) and Te(r) profiles and study transport and other fundamental aspects of the discharge.

In addition to the more familiar plasma physics and fusion related experiments, the ECH system is also being used in an experiment in conjunction with NASA as the power source for small rockets being accelerated by microwave heating of fuel without use of an oxidizer. Tests of the design of suitable heat exchangers and engines are ongoing.

The long-term upgrade plan for the DIII-D gyrotron complex calls for 10 gyrotrons with a regular replacement program leading to most or all of the tubes at 117.5 GHz with nominal generated power of 1.5 MW per unit. This will require upgrades to the building, new high voltage power supplies, one additional dual launcher, and controls to expand beyond the present 8 gyrotron system.

This work was supported by the US Department of Energy under DE-FC02-04ER54698 and DE-AC02-09CH11466.

Machine safety issues with respect to the extension of ECRH systems at ASDEX Upgrade

M. Schubert, A. Herrmann, F. Monaco, H. Schütz, J. Stober, T. Vierle, S. Vorbrugg,D. Wagner, D. Zasche, T. Zehetbauer, W. Zeidner and ASDEX Upgrade Team

Max Planck Institute for Plasma Physics, EURATOM Association,Boltzmannstr. 2, 85748 Garching, Germany

Although in general the absorption of ECRH in fusion plasmas is rather high, there is always a small fraction of non-absorbed power in the order of a few percent. This so-called stray radiation may damage in-vessel components. If a particular heating scheme fails, the non-absorbed fraction may increase dramatically. Even a direct hit of the compact, high-power microwave beam on plasma facing components is possible.In the vicinity of the ECRH launchers at ASDEX Upgrade, damages are reported during regular inspections. Two major categories may be classified: On one hand there is the continuous degradation of absorbing, insulating material. Often this material is not even facing the plasma. The degradation is probably due to the absorption of millimeter wave stray radiation that entered the cavities behind the plasma facing tiles. On the other hand, local damages, even of metallic components, were observed on surfaces which were exposed directly to the microwave beam. These damages are attributed to singular hazardous events.Protection measures are carried out in order to prepare upcoming experimental campaigns. In-vessel components are hardened against stray radiation. This holds in particular for the inner heat shield. Gaps between tiles are replaced by a labyrinth. In the cavities between the plasma facing tiles and the wall of the vacuum vessel, actively cooled metallic surfaces have been coated with a layer of titanium dioxide and aluminum oxide. The layer should act as an absorber for the stray radiation that enters these cavities.As to the prevention and mitigation of single hazardous events, two actions are taken. Firstly, polarization control is automated. The actual mode content is calculated using magnetic configuration data, polarizer and launcher settings, and the result is compared to the nominal heating scheme. In this way it should be possible to intercept operator errors. Secondly, the number of channels for the real-time monitoring of the stray radiation level is increased. Since the stray radiation maxima can be very localized, this seems to be important particularly in the toroidal segments adjacent to the locations of the ECRH launchers.

Recent upgrading of ECRH system and studies to improve ECRH performance in the LHD

H. Igami, S. Kubo, T. Shimozuma, Y. Yoshimura, H. Takahashi, S. Kobayashi, S. Ito, Y. Mizuno, K. Okada, *R. Makino, *S. Ogasawara, *K. Kobayashi, M. Osakabe, **K. Nagasaki, ***H. Idei, T. Mutoh, and the LHD experimental group National Institute for Fusion Science, Toki, 509-5292, Japan *Department of Energy Engineering and Science, Nagoya Univ. Nagoya, 464-8603, Japan **Institute of Advanced Energy, Kyoto Univ. 611-0011 Uji, Japan, ***Research Institute for Applied Mechanics, Kyushu Univ. 816-8580 Kasuga, Japan e-mail: igami@LHD.nifs.ac.jp In the latest experimental campaign of the large helical device (LHD), three 77GHz/~1MW/5s and one 154GHz/~1MW/5s Triode CPD gyorotrons were mainly operated for electron cyclotron resonance heating (ECRH) in short pulse discharges. These gyrotrons were developed under collaboration with Tsukuba Univ. and fabricated by TOSHIBA. With power enhancement method by stepped anode acceleration voltage control about 4.5 MW total simultaneous injected power has been achieved with adding the operation of one 82.7GHz GYCOM gyrotron. Enhancement of the injected ECRH power allowed us to achieve Te0 ~ Ti0 ~ 6keV operation, increase of electron temperature in over-dense plasmas by excitation of the electron Bernstein wave, control the current profile to prevent the electron temperature flattening with maintaining the electron internal transport barrier (e-ITB) by adequate electron cyclotron current drive (ECCD) and so on. For long pulse operation, with alternately switched operation of two 77GHz gyrotrons at 155kW and 110kW and continuous operation of a 84GHz gyrotron at 110kW, a plasma of

€

˜ n e < 1.0x1019m-3 and Te0 ~ 1.5keV was maintained for 30 minutes by 0.24MW power injection in total. To improve the ECRH performance more, it is important to increase the absorption power by optimization of the launching condition. Recently, for frequently used settings of the wave injection angle, it has been derived that the 77GHz ECRH efficiency estimated by change of the time derivative of the stored energy between the start/end of ECRH is about 20% less than that of 154GHz ECRH in low density (

€

˜ n e < 1.0x1019m-3) discharges. As a possible reason, it is pointed out that the wave power that couples with the heating modes (fundamental ordinary (O-) mode or the second harmonic extraordinary (X-) mode) deep into the plasma is reduced by mode coupling effect between the O- and the X-modes caused by magnetic shear outside the last closed flux surface (LCFS). For the case of the second harmonic ECRH by 77GHz EC wave, it has been found that a high ECRH efficiency is not necessarily obtained by right-handed circularly polarized (R-circ.) wave injection although the injected wave has sufficiently large |N//| at the LCFS. In the LHD, the ergodic layer where the low density plasma is weakly confined expands widely outside the LCFS along the long axis direction of the poloidal cross-section because of the existence of the intrinsic divertor. Since the poloidal cross section rotates along the toroidal direction in the LHD, the transverse distance of the injected beam in the ergodic layer changes as the toroidal injection angle changes. Reduce of the ECRH efficiency is observed when the transverse distance is long. The polarization of the injected wave can change along the propagation before the wave reaches the LCFS by mode coupling effect. It is required to adjust the polarization of the injected wave to couple with the X-mode inside the LCFS efficiently with taking into account the mode coupling effect for more efficient ECRH/ECCD.

Control of Energetic-Particle-Driven MHD Modes by ECH/ECCD in Helical Systems

K. Nagasaki1, S. Yamamoto1, S. Kobayashi1, K. Nagaoka2, E. Ascasibar3, M. Osakabe2, Y.

Yoshimura2, T. Mizuuchi1, H. Okada1, T. Minami1, S. Kado1, S. Ohshima1, S. Konoshima1, N. Shi1, K. Sakamoto1, Y. Nakamura4, L. Zang4, N. Kenmochi4, F. Volpe5, N. Marushchenko6, F.

Sano1 and the LHD experiment group

1Institute of Advanced Energy, Kyoto University, Kyoto, Japan 2National Institute for Fusion Science, Toki, Japan

3CIEMAT, Madrid, Spain 4Graduate School of Energy Science, Kyoto University, Japan

5Columbia University, New York, USA 6Max-Planck-Institut für Plasmaphysik, EURATOM, Teilinstitut Greifswald, Germany

E-mail address: nagasaki.kazunobu.4x@kyoto-u.ac.jp

Understanding and control of magnetohydrodynamic (MHD) modes is an important issue

for sustaining high-perfomance plasmas in toroidal fusion devices such as tokamaks and helical systems. Energetic particles (EP) can interact resonantly with shear Alfvén waves during slowing-down process, and excite MHD instabilities, that is, Alfvén eigenmodes (AEs), resulting in enhanced radial transport of the energetic ions and serious damage of plasma facing components. ECH/ECCD is an ideal tool to suppress EP driven MHD modes since it can provide highly localized EC current with a known location and good controllability, which enables to change pressure profile and to tailor rotational transform profile (q profile) and its shear. Effects of ECH/ECCD on the MHD modes were investigated in several tokamaks and helical systems. In TJ-II, reduction of the AE amplitude was observed when the ECH power was applied, and the continuous character of the unstable AEs changed to a chirping mode marginally unstable [1].

In this paper, we show recent exprimental results on control of EP-driven MHD modes by ECH/ECCD in the Heliotron J and LHD devices. In Heliotron J, energetic particle modes (EPM) excited by neutral beams have been successfully stabilized by applying second harmonic 70-GHz X-mode ECCD [2]. The EPM of 80-100 kHz are stabilized by both co- and counter-ECCD of a few kA which forms a magnetic shear. No stabilization effect is observed for ECH only in the same magnetic configuration. These experimental results indicate that both positive and negative magnetic shear are effective for EPM stabilization. In LHD, an AE of 260 kHz excited in an NBI plasma has been stabilized by overlapping second harmonic X-mode 77-GHz ECH, while another mode of 210 kHz has been excited at the same time. Stabilization of an AE by co-ECCD has been also observed. The physical mechanism of mode excitation and stabilization by ECH/ECCD will be discussed. [1] K. Nagaoka, et al., Nucl. Fusion 53 (2013) 072004 [2] K. Nagasaki, et al., Nucl. Fusion 53 (2013) 113041

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Fig2. OH target plasma with

8.2 + 28 GHz injection

resulting high βp and nullformation. Visible cameraimage clearly showsseparatrix strike points oninboard limiter withoverlaid magnetic fluxcontours shown.

time= 2.4 s

Fig1. Typical OH plasma with 8.2 GHz

ECCD. Rin >0.22 m depicts the null formation

at βp* > 3. Magnetic flux contours shows

inboard limiter to natural divertor transition.

LCFS calculated through analytic model is

shown with that of the magnetic measurement.

Fig2. OH+ 8.2+

28GHz ECCD

plasma. Visible

camera image

shows separatrix

strike points on

inboard limiter.

High βp plasma formation using off-axis ECCD in Ohmic heated plasma in

the spherical tokamak QUEST.

Kishore Mishraa, H. Zushi

b, H. Idei

b, M. Hasegawa

b, K. Hanada

b and QUEST team

a IGSES,

b RIAM, Kyushu University, Kasuga, Fukuoka, Japan, 816-8580

mishra@triam.kyushu-u.ac.jp

High poloidal beta (βp ~1) operation in steady

state condition in tokamaks is of great interest and has

previously been demonstrated using NBI, LHCD and

low current (Ip) plasma [1-2] for a short time (<0.5 s).

A very few experiments however, have been

performed towards the investigation of highest

obtainable βp in tokamak plasma. In this work we

report the first result of high βp production and its

long sustainment though an off axis ECCD at two

different frequencies in Ohmic (OH) target plasma.

OH plasma is formed at fixed Ip with the help of

an external feedback circuit and ECW is injected into

the flattop of Ip. Two low field side phased waveguide

array antennae are employed to inject 8.2 GHz ECW

at O-mode corresponding to the fundamental

resonance Rfce =0.32 m (R0=0.68 m). A highly curved

vertical magnetic field with Bz/Bt ~ 0.1 is applied at

Rfce to confine the energetic electrons produced

through multiple resonant interaction of ECW[3],

which induces a transient increase in the Ip and

recharging of OH circuit. Owing to better confinement

of the energetic electrons[4], non-thermal pressure in

plasma is enhanced and a βp* (

= βp + li / 2) > 4 is

achieved. As βp* increases beyond a critical value of ~

3, the plasma configuration is changed from inboard

limiter (OH phase) to naturally create an inboard

poloidal field null indicating that plasma equilibrium

has encountered a βp limit (βp = 1+/2). The formation

of such null point is independently confirmed from the

shift in density centroid measured by Thomson

Scattering and measurement of its separatrix strike

points on the inboard limiter through a visible camera.

With the transition in the configuration, the kink safety

factor (q*) is reduced from above 4 to below 2, however, the equilibrium is found stable for >

1s without any observable MHD instabilities. Similar experiments are carried out at second

harmonic ECCD using 28GHz Gyrotron with R2fce =0.32 m in OH target plasma with

simultaneous 8.2 GHz injection at Rfce =0.54 m . In this scenario, high βp plasma is formed

with appearance of an inboard null as observed in the earlier occasion. However, reduction in

q* is not observed in this case and rather an increment from 5 to 6 is observed. The two

equilibrium scenarios in ECCD, namely off-axis fundamental and off-axis second harmonic

in presence of an on-axis fundamental resonance are investigated and will be reported. [1] M. E. Mauel et. al., Nuc. Fusion 32, 1468 (1992) [2] S. C. Luckhardt et. al., Phy. Rev. Lett. 62, 1508 (1989)

[3] H. Zushi et. al., IAEA, EX/P2-14 (2012) [4] S. Tashima et. al. accepted for Nuclear Fusion, (2014)

Non-inductive Current Drive Experiments using 28 GHz Electron Cyclotron Waves in QUEST

H. Idei, *T. Kariya, H. Zushi, K. Hanada, *T. Imai, O. Watanabe, **K. Mishra and QUEST team Research Institute for Applied Mechanics, Kyushu University. Kasuga 816-8580, Japan * Plasma Research Center, University of Tsukuba, 305-8577, Tsukuba, Japan ** IGES, Kyushu University. Kasuga 816-8580, Japan e-mail: idei@kyushu-u.ac.jp 2nd harmonic 28 GHz Off-axis ECCD Maximum toroidal field has been B0 = 0.25 T at a major radius R = 0.64 m using a Toroidal Field Coil (TFC) power supply with its maximum current of 50 kA in the QUEST. 2nd harmonic Electron Cyclotron Resonance (ECR) layer was located inside plasma at R = 0. 32 m due to the low aspect configuration. First the 28 GHz ECCD effect was confirmed in ohmically sustained plasma of 30 kA with feedback regulation of CS coil current. Figures1 show time evolutions of TFC current, vertical field Bv, CS coil current and plasma current Ip with and without the superposed 28 GHz injection. Sine the ohmic discharges have been conducted with rather low toroidal field, the plasma was started up at the low field and the toroidal field was ramped up to the maximum. Although the plasma current has been controlled with the feedback, it began to increase by the 28GHz injection against retarding electric field by ramp-up of the CS coil current. Recharging phenomena in the CS coil power supply were clearly observed. The plasma current was fully driven only by the 2nd harmonic off-axis 28GHz injection. 28 GHz Plasma-Sustainment in Low Aspect Configuration Fully non-inductive current drive experiments have been conducted with the 28 GHz ECH/ECCD system in the QUEST. Figures 2 show time evolutions of plasma current Ip, vertical field Bv, loop voltage, plasma shaping parameters of minor radius a, elongation and triangularity in the non-inductive limiter-configuration plasma. The plasma was pre-ionized by the 4 kW 8.2 GHz injection. The plasma current was rapidly ramped up against retarding loop voltage due to the current ramp-up itself. The observed current ramp-up rate was 0.5 MA/s. The plasma current was finally ramped up to 54 kA in accordance with the vertical field, and was sustained for 0.9 sec. The plasma shaping was almost kept for 1.3 sec. The magnetic axis radius and the aspect ratio were 0.67 m and 1.4, respectively. In Bv ramp-up experiments, the plasma current of 66 kA was non-inductively attained with the 28 GHz injection. Hard X-ray intensities with more than 100 keV energy were measured in the non-inductive 28 GHz plasma.

Figs.1: Time evolutions of TFC coil current, vertical field Bv, the CS coil current and plasma current Ip w/ and w/o superposed 28GHz injections.

Figs.2: Time evolutions of plasma current Ip , vertical field Bv , loop voltage, plasma shaping parameters of minor radius a, elongation κ and triangularity δ in the non-inductive limiter-configuration plasma.

Operation regime of RT-1 by electron cyclotron heating M. Nishiura, Z. Yoshida, H. Saitoh, Y. Yano, Y. Kawazura, T. Nogami, M. Yamasaki Graduate School of Frontier Sciences, The University of Tokyo, Chiba 277-8561, Japan e-mail address : nishiura@ppl.k.u-tokyo.ac.jp

A magnetosphere configuration produced by a dipole magnet is one of the concepts for thermonuclear fusion reactor [1]. The ring trap 1 (RT-1) based on this concept is equipped with a superconducting coil in a vacuum to produce a dipole field for plasma confinement. The magnetic configuration is attractive to realize high performance plasma which leads to a fusion reactor as well as to a prediction of the nature of plasmas in magnetosphere.

The RT-1 device produces high beta plasmas by an electron cyclotron heating (ECH) system with 2.45 GHz and 8.2 GHz microwaves. The operation regime has been extended to a higher electron density and beta regime by an increase of ECH power up to 40 kW with an 8.2 GHz klystron. At 40 kW, the diamagnetic flux Wp and line integrated electron density n increase up to about 1.6 times compared with those at 25 kW in the levitated coil configuration. However, when the dipole magnet is not levitated, Wp starts to saturate around 10 kW. By levitating the magnet and reducing the interactions between the plasma and the coil support structure, Wp increases about 10 times. When the filling gas pressure is increased from 3.8 mPa to 13.0 mPa, n reaches ~ 5×1017 m-3 by the core chord of the interferometer.

The high beta plasma in a dipole magnetic field is characterized by a strongly peaked density profile [2], which is explained by a kinetic equilibrium theory [1, 3]. The peaked density profile is produced by the inward (or, up-hill) particle diffusion [4]. The above high density regime would be expected to be a peaked density profile. As a complementary approach to the experimental evidence of the density peaking, a 4.2 GHz reflectometer has been installed to measure the local position of the cut off density. The initial result shows that the reflected signal was observed at a high density operation. However we need careful analysis to prove the formation of the cut off layer by a density peaking.

At more than 40 kW injection, as the electron cyclotron resonance layer for the fundamental frequency exists in an over dense plasma, if the peaked density is created. In such plasmas, the heating efficiency by ECH and the characteristics on produced plasmas are compared experimentally with X-mode and O-mode injections. The result shows the achievable Wp and n were increased up to 5.6 mWb and 6.2×1017 m-3 in the case of simultaneous X-mode and O-mode injections from separated two ports, compared with only in the case of O-mode injection. The ray trace calculation would support the optimized injection scheme to extend the plasma parameters in RT-1. For the ion heating, an ion cyclotron heating (ICH) system is implemented in RT-1. The optimum density for ICH is more than ~ 1018 m-3 from a code prediction. The present ECH system achieved the sufficient electron density as target plasmas for ion heating.

References [1] A. Hasegawa et al., Nucl. Fusion 30 (1990) 2405. [2] H. Saitoh et al., Nucl. Fusion 51 (2011) 063034. [3] Z. Yoshida et al., Plasma Phys. Control. Fusion 55 (2013) 014018. [4] Z. Yoshida et al., Phys. Rev. Lett. 104 (2010) 235004.

Status of the Design of the ITER ECE Diagnostic G. Taylor, R. Feder, D. W. Johnson, C. Roman

Princeton Plasma Physics Laboratory, Princeton, NJ 08543, USA H. K. B. Pandya, S. Danani, R. Kumar, S. Kumar

ITER-India/Institute for Plasma Research, Bhat 382428, Gandhinagar, India M. E. Austin, P. E. Phillips, W. L. Rowan

Institute for Fusion Studies, University of Texas at Austin, TX 78712, USA J. H. Beno and A. Ouroua

Center for Electromechanics, University of Texas at Austin, TX 78758, USA R. F. Ellis

University of Maryland, College Park, MD 20742, USA A. E. Hubbard

Plasma Science and Fusion Center, MIT, Cambridge, MA 02139, USA V. Udintsev, G. Vayakis, M. Walsh

ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance, France

The baseline design for the ITER electron cyclotron emission (ECE) diagnostic system has entered the detailed preliminary design phase that will continue until the preliminary system design review in October 2015. Two plasma views are currently being planned, a radial view and an oblique view that is sensitive to distortions in the electron momentum distribution near the average thermal momentum. Both views can provide high spatial resolution temperature profiles if the electron momentum distribution remains Maxwellian. The ECE diagnostic system can be divided into three main parts; 1) the front-end optics, including the in-situ 1000 K calibration sources, in Equatorial Port Plug 9 (EP9), 2) the polarization splitter box and broadband transmission system that transports the ECE from the front-end and distributes it to the Diagnostics Hall, and 3) the ECE instrumentation in the Diagnostics Hall. Review of the final design is expected in August 2016 and will be split into several final design reviews covering components in EP9, the Port Cell and Galleries, and the Diagnostics Hall. The baseline ECE instrumentation will include a two channel Michelson interferometer system that will simultaneously measure ordinary and extraordinary mode polarized ECE over a broad spectral range from 70 GHz to 1 THz, and two heterodyne radiometers, covering 122-230 GHz and 244-355 GHz. The US Domestic Agency (US-DA) will provide the front-end components, the 244-355 GHz radiometer and the control cubicle in the Diagnostics Hall that will control and monitor the calibration sources and shutters in the front-end. The India Domestic Agency (IN-DA) will provide the remaining components of the ECE system. Control, data acquisition and analysis software will be developed jointly by the US-DA and the IN-DA. The US-DA is responsible for integrating the front-end of ECE system into EP9 and the ITER organization is responsible for external interfaces, for example with buildings. Significant design challenges include 1) developing reliable 1000 K hot calibration sources and the associated remotely controlled shutters/mirrors in EP9, 2) providing compliant couplings in the millimeter wave transmission lines between the front-end optics and the polarization splitter box that can accommodate displacements of the vacuum vessel associated with thermal expansion during plasma operations and bake out, and during plasma disruptions, 3) protecting the ECE diagnostic components from stray ECH radiation and other intense millimeter wave emission and 4) providing a low-loss broadband transmission system between the front-end and the Diagnostics Hall. This presentation will review the current design of the baseline ECE diagnostic, design challenges that will need to be addressed during the preliminary and final design phases, and possible future upgrades to state-of-the-art ECE instrumentation.

New approach to ECE measurements based on Hilbert-transform

spectral analysis

Hitesh Kumar B. Pandya

ITER-India, Institute for Plasma Research, Gandhinagar-380025, India; hitesh@ipr.res.in

and

Yuriy Divin

Peter Grünberg Institute, Forschungszentrum Jülich, 52425 Jülich, Germany ; Y.Divin@fz-juelich.de

Abstract:

Spectroscopy of Electron Cyclotron Emission (ECE) has been established as adequate

diagnostic technique for fusion research machines. Among various instruments for ECE

diagnostics, only Fourier-transform spectrometers with Martin-Puplett Interferometers measure

electron cyclotron radiation in a broadband frequency range from 70 to 1000 GHz. Before these

measurements, total system including a front-end radiation collector, a transmission line, an

interferometer and a radiation detector should be absolutely calibrated. A hot/cold calibration

source and data-averaging technique are used to calibrate the total ECE diagnostic system. It

takes long time to calibrate the ECE system because of the low power level of the calibration

source and high values of the noise equivalent power (NEP) of the detection system. A new

technique, Hilbert-transform spectral analysis, is proposed for the ITER plasma ECE spectral

measurements. An operation principle, characteristics and advantages of the corresponding

Hilbert-transform spectrum analyzer (HTSA) based on a high-Tc Josephson detector are

discussed. Due to lower NEP-values of the Josephson detector, this spectrum analyzer might

demonstrate shorter calibration times than that for the Martin-Puplett Interferometer.

Key words: Hilbert transform spectrometer, Electron Cyclotron Emission (ECE), Fourier-

transform spectrometer (FTS)

EC Radiative Transport and Losses in DEMO-like High-Temperature Plasmas F. Albajar1, M. Bornatici2 and F. Engelmann3

1 Fusion for Energy, Josep Pla 2, Barcelona, 08019, Spain 2 Physics Department, University of Pavia, Pavia, 27100, Italy 3 Max-Planck-Institut für Plasmaphysik, Garching, 85748, Germany In view of the potential benefits of operating DEMO at high plasma temperature (1), EC radiative transport and losses are of particular relevance since in this case the locally radiated EC power density in the plasma core tends to become significant with respect to the power density of fusion α-particle heating (2,3). Here, using the RAYTEC code (4) complemented to cover α-particle heating using Brunelli’s fit (5) for the DT fusion cross-section, a detailed analysis of EC radiation in high-temperature DEMO-like plasmas is presented. Specifically, parametric studies of the radial EC radiation profile as well as the total power loss have been carried out in the electron temperature (central values in the range 20 to 80 keV; “advanced” and parabola-like profiles (4)), in the toroidal magnetic field (B = 6, 8, 10, 12 T) and in the wall reflection coefficient (RW = 0.6 and 0.8), combined with various choices of the plasma density and density profile (a flat (4) and a peaked profile as suggested in Ref.(6)). The results obtained are in line with, and complement earlier findings attained with global models. In particular, it is displayed again that the relative importance of EC radiation effects, for equivalent plasma conditions, is independent of the value of the confining toroidal magnetic field.

(1) D. J. Ward, Plasma Phys.Control.Fusion 52 (2010) 124033 (2) F. Albajar, M. Bornatici, G. Cortes, et al., 2005 Nucl. Fusion 45 (2005) 642 (3) F. Albajar, M. Bornatici and F. Engelmann, 2nd IAEA DEMO Programme Workshop,

17 -20 December 2013, Vienna, Austria (4) F. Albajar, M. Bornatici and F.Engelmann, Nucl.Fusion 49 (2009) 115017 (5) B. Brunelli, Il Nuovo Cimento 55B (1980) 264 (6) H. Zohm, C. Angioni, E. Fable, et al., Nucl.Fusion 53 (2013) 073019

The effect of synchrotron radiation loss on conductivity of

a relativistic magnetized plasma

Y. R. Lin-Liu

Department of Physics and Center for Mathematics and Theoretical Physics, National Central University, Taiwan

Abstract: It is well known that synchrotron radiation loss could play an important role in energy balance of a tokamak reactor.[1] A passive current drive scheme utilizing synchrotron radiation for current sustainment in tokamak reactors, especially for those using advanced fuels, was a subject of intensive investigation some years ago.[2-5] On the other hand, the effects of synchrotron radiation loss on conductivity and current drive efficiency of externally imposed current drives have not been systematically investigated for hot magnetized plasmas. It is generally believed that the effects are of little significance for ITER plasmas, but the situation becomes less clear for higher temperature plasmas, such as the DEMO plasmas. The purpose of this work is to provide some quantitative estimates of an upper bound of the effect for the problem of conductivity. The radiation loss is represented by the friction-force model of Bernstein and Baxter[6] and is treated as a perturbation to Coulomb collisions, which are modeled by the weakly relativistic collision operators[7]. Corrections to the Ohmic conductivity due to the radiation loss are expressed as a function of electron plasma beta, electron temperature, and the effective ion charge. Numerical results of the parameter regime relevant to the DEMO reactor will be presented. [1] B. A. Trubnikov, in Reviews of Plasma Physics, edited by M. A. Leontovich (Consultants Bureau, New York, 1979), Vol. 7 p.345. [2] J. M. Dawson and P. K. Kaw, Phys. Rev. Lett. 48, 1730 (1982). [3] I. Fidone, G. Granata, and J. Johner, Phys. Fluids 31, 2300 (1988). [4] I. Fidone, Phys. Fluids B5, 2300 (1993). [5] S. V. Kasilov and W. Kernbichler, Phys. Plasmas 3, 4115 (1996). [6] I. B. Bernstein and D. C. Baxter, Phys. Fluids 24, 108 (1981). [7] C. F. F. Karney and N. J. Fisch, Phys. Fluids 28, 116 (1986). Session: theory, poster

Development of Momentum Conserving Monte Carlo Simulation Code

for ECCD Study in Helical Plasmas

S. Murakami, S. Hasegawa and Y. Moriya

Department of Nuclear Engineering, Kyoto University, Nishikyo, Kyoto 615-8530, Japan

The conservation of the momentum during particle collisions is an important issue in studying

the electron cyclotron current drive (ECCD), the neoclassical transport and etc. In this study

the velocity dependent model is derived from the Fokker–Planck collision term directly and

is implemented to global Monte Carlo simulation code GNET[1], in which the linearized drift

kinetic equation for energetic particles distribution, δ f (x,v‖,v⊥, t) = f (x,v‖,v⊥, t)− fmax(r,v2),

is solved in 5-D phase space, introducing an iterative process.

In order to conserve the momentum, we assume the particle collision term of δ f (x,v‖,v⊥, t)

as Ccoll(δ f ) =C(δ f , fmax)+C( fmax,δ f ), where C( fmax,δ f ) is the field particle operator which

represents the collision effect for the background particles. The field particle operator can be ex-

pressed using Legendre polynominals and, introducing the Trubnikov-Rosenbluth potential, we

can derive the field particle term for each Legendre polinominals, Cn( fmax,δ f (n)(v)). Once the

field particle operator is obtained we can consider C( fmax,δ f ) as a new source term in the drift

kinetic equation. In the GNET code, we introduce an iterative process to implement the mo-

mentum conserving collision operator. We first obtain the steady state solution, δ f0, assuming

an original source term, Sql and evaluate C( fmax,δ f0), which becomes a new source term. Next

we obtain a steady state solution, δ f1, with this new source term and evaluate C( fmax,δ f1).

Then we iteratively obtain δ fn until the lost momentum close to zero.

The developed model is applied to the ECCD simulation of the heliotron-J[2]. The simulation

results show a good conservation of the momentum and the increment of ECCD current is

observed.

References

[1] S. Murakami et al., Nucl. Fusion 40, 693 (2000).

[2] Y. Moriya et al., Plasma Fusion Res. 6, 2403139 (2011).

Assessment of the ITER EC Upper Launcher Performance

L. Figini1, D. Farina1, E. Poli2, A. Bruschi1, A. Moro1, P. Platania1, C. Sozzi1, T. Goodman3, O. Sauter3, M. Cavinato4, G. Saibene4, M.A. Henderson5

1Istituto di Fisica del Plasma CNR, Euratom-ENEA Association, 20125 Milano, Italy 2Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching, Germany 3EPFL-CRPP, Association EURATOM, Confédération Suisse, CH-1015 Lausanne, Switzerland

4Fusion for Energy, C/ Josep Pla no 2, 08019 Barcelona, Spain 5ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance, France

The ITER Electron Cyclotron (EC) Heating and Current Drive (H & CD) system, operating at 170 GHz, will deliver up to 24 MW to the plasma through one Equatorial (EL) and four Upper Launchers (UL) [1]. The capabilities foreseen for the system include central heating and bulk current drive, the control of Magneto-Hydrodynamic (MHD) activity such as Neoclassical Tearing Modes (NTMs) at the q=3/2 and q=2 surfaces and sawtooth instabilities at the q=1 surface, the assist to break-down and L- to H-mode transition phases of the plasma discharge. Moreover, it is required to be effective both when ITER will operate at nominal and reduced magnetic field magnitude [2, 3]. The functionalities have been partitioned between the two types of launchers, assigning the task of central heating and bulk current drive mainly to the EL, and the control of NTMs to the UL, that has consequently been designed to inject focused beams – four on each steerable launching mirror – in the outer half of the plasma radius. This paper presents the analysis performed so far to assess the performance of the UL, and the requirements on its design to achieve all the required functionalities. Different scenarios have been studied, along the full temporal evolution of the discharge, including the reference ITER 15MA H-mode plasma, a case with reduced magnetic field strength at 2.65T, and a steady state scenario. The steering range necessary to reach the q=3/2 and q=2 surfaces from both the Upper (USM) and the Lower Steering Mirror (LSM) have been determined for the different scenarios. By making use of two different criteria [4, 5], each relevant in the opposite limits of small and large width of the driven current channel, and by a direct solution of the Rutherford equation [6], the optimal launching angles and the power required for NTMs stabilization has been estimated and found to be sufficient in the main phases of the discharge. The increase in power requirements due to partial misalignment among the beams launched from the same mirror has also been estimated. These results, together with the engineering limits and the additional functionalities that the launcher must achieve, will be used to drive the optimization and finalization of the UL design. References

[1] M.A. Henderson et al., Nuclear Fusion, 48 054013 (2008). [2] D. Farina et al., Nuclear Fusion, 52 033005 (2012). [3] D. Farina et al., Physics of Plasmas, accepted (2014). [4] H. Zohm et al., Plasma Physics and Controlled Fusion, 49 B341 (2007). [5] O. Sauter et al., Plasma Physics and Controlled Fusion, 52 025002 (2010). [6] N. Bertelli et al., Nuclear Fusion, 51 103007 (2011).

A Megawatt-Level 28 GHz Heating System for the National Spherical Torus Experiment Upgrade*

G. Taylor, R.A. Ellis, E. Fredd, S. P. Gerhardt, N. Greenough, J. C. Hosea, F. Poli Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543, USA

R. Parker, S. Shiraiwa, D. Terry, G. Wallace, S. Wukitch Plasma Science and Fusion Center, MIT, Cambridge, MA 02139, USA

R. Raman University of Washington, Seattle, WA 98195, USA

R. W. Harvey CompX, Del Mar, CA 92014, USA

A. P. Smirnov M.V. Lomonosov Moscow State University, Moscow, Russia

Construction of the National Spherical Torus Experiment Upgrade (NSTX-U) [1] is currently being completed and first plasma operations are expected in late 2014 or early 2015. NSTX-U will operate at axial toroidal fields of up to 1 T and plasma currents (Ip) up to 2 MA. The development of fully non-inductive (NI) plasmas is a major long-term research goal for NSTX-U that supports the design of a Fusion Nuclear Science Facility [2]. A megawatt-level, 28 GHz electron heating system is being designed to heat NI NSTX-U start-up plasmas in 2017-18. This heating system will later be used to provide radially localized electron heating and current drive during the Ip flat top of overdense H-mode discharges [3]. 0.6 MW of 28 GHz electron cyclotron (EC) heating is expected to increase the central electron temperature (Te(0)) of low density NI plasmas generated by Coaxial Helicity Injection (CHI) [4] in NSTX-U from 10 eV to 400 eV in about 20 ms. The increased Te(0) will significantly reduce the Ip decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. The 28 GHz heating system will also be used for electron Bernstein wave (EBW) plasma start-up using a technique developed on MAST [5]. The present pre-conceptual design of the 28 GHz heating system plans to use a gyrotron designed at the Tsukuba University Plasma Research Center [6]. The gyrotron will be a depressed collector upgrade of an existing 1 MW unit and is being designed to extend the maximum power to 1.5-2 MW. An existing modular insulated-gate bipolar transistor power supply will be used as the main power supply for the gyrotron. The supply is capable of serving as the main power supply for two gyrotrons. This paper will present the latest design for the NSTX-U 28 GHz heating system and numerical simulation results for 28 GHz EC and EBW heating and current drive for NSTX-U discharges. [1] J. E. Menard, et al., Nucl. Fusion 52, 083015 (2012). [2] Y.-K.M. Peng, et al., Fusion Sci. Technol. 56, 957 (2009). [3] S.P. Gerhardt, et al, Nuclear Fusion 52, 083020 (2012). [4] R. Raman, et al, Nucl. Fusion 53, 073017 (2013). [5] V. F. Shevchenko et al., Nucl. Fusion 50, 022004 (2010). [6] R. Minami, et al, Nucl. Fusion 53, 063003 (2013). *Work supported by USDOE Contract No. DE-AC02-09CH11466

Polarization and mode control of EAST 140 GHz ECH system

Dajun Wu1, Fukun Liu

1, Jiafang Shan

1, Handong Xu

1, Xiaojie Wang

1, Wei Wei

1,2,

Yunying Tang1

1Institute of plasma physics, Chinese Academy of sciences, Hefei 230031, China

2HeFei University of Technology, Hefei, China

The 140 GHz electron cyclotron heating and current driving systems are under

construction on EAST tokamak, which designed to launch 4MW of power for the

duration up to 100s into the plasma. The heating and current driving efficiencies

depend on the wave mode in plasmas, so polarization control is necessary for

effective plasma heating and current driving on the ECRH/ECCD system. Two

polarizer miters will be used to control the wave polarization and ellipticity for each

transmission line on EAST ECRH/ECCD system, any wave polarization and

ellipticity can be produced by changing the mirror rotation angle of each polarizer

miter bend. This paper shows the required polarizer mirror settings as a function of

the injection angles for pure second extraordinary harmonic mode.

The EAST 140 GHz 4 MW ECRH system

Xiaojie Wang1, Fukun Liu

1, Jiafang Shan

1, Handong Xu

1, Dajun Wu

1, Bo Li

1, Wei

Wei12

, Yunying Tang1, Weiye Xu

1, Huaichuan Hu

1, Jiang Wang

1, Li Xu

1

1Institute of Plasma Physics Chinese Academy of Sciences, Hefei, China

2Hefei University of Technology, Hefei, China

A 140 GHz electron cyclotron heating and current drive (ECH &CD) system for

EAST Tokamak is under development. The system is designed to inject 4 MW power

with the duration up to 100 s. The main objectives of the system are to provide central

heating, off-axis current drive, assist start-up and control of MHD activities. The

system has four gyrotrons each with norminal 1MW output power and 1000s pulse

length. Each gyrotron is connected to the tokamak by a low-loss evacuated waveguide

transmission line. Four beam lines will be injected into the plasma through an actively

cooled launcher located in an equatorial port. The front steering launcher can direct

the RF beam over ±25° toroidally and scan over 30° poloidally. In this paper, a brief

overview and the development of the 140-GHz ECRH system are presented.

Session: ECRH/ECCD

Real-Time Feed-backed Anode Power System for 170GHz gyrotron in KSTAR

1Bong-jun Seok, 1Il-kum Ahn,1Seung-kyo Lee,,1Eun-yong Shim,2Young-soon Bae,2Jin-hyun Joung, 2Mi-Joung, 2Won-soon Han

1Dawonsys company,227, Gyeonggigwagidae-ro, Siheung-si, Gyeonggi-do 429-850, Korea

2National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon 305-333, Korea

A 3.6 MW (66 kV/55 A) high voltage power supply system was developed for the 170 GHz ECH&CD system

in KSTAR. It consists of the cathode power supply (CPS), anode power supply (APS), and body power supply

(BPS).

Currently, modulation frequency of APS was limited to less than 1 kHz due to the parasitic capacitance of

Zener diode circuit including gyrotron. The APS consists of fixed resistors and many series connections of

Zener diodes. Each Zener diode has the voltage holding of 200V. The input of APS is connected to the output of

CPS with the reference of the ground potential. The cathode-anode voltage is not actively regulated following

the fluctuation of the cathode voltage. In order to solve these lacks of controllability and high frequency

modulation of APS, the existing APS is replaced by a new type of APS and it is tested for KSTAR 170 GHz

gyrotron system.

The new APS is composed of a Cap Charger Power Supply (CCPS), High Voltage Switch (HVS) Tank. The

CCPS has specification of rated energy at maximum 35kJ/S and the control frequency at 20 kHz. The storage

energy of the CCPS is supplied to cathode-anode of gyrotron pursuant to HVS tank. The HVS consists of a

on/off switch and a discharging switch which are made by parallel and series connections of many MOSFET

fast solid-state switch of which switching on-off time is less than 1 microsecond.

This configuration reduced significantly the total parasitic capacitance between cathode and anode in gyrotron.

The developed new APS has specification of rated voltage at 0~50 kV, rated current at maximum 1 A. Also, the

voltage output is regulated with stability of 0.3% and the modulation is possible to up to 10 kHz with the rising

and falling time less than ten microseconds.

The output of CCPS is rapidly feedback controlled by real time control board using analog comparator circuit.

Since the CCPS reference (common to chassis) is connected to the CPS output, the cathode-anode voltage will

be supplied with constant level. In addition, the active cathode-anode voltage control capability will be

applicable for the active beam current control to overcome the beam emission cooling effect for the long pulse

operation of the gyrotron

This paper presents the new APS design and characteristics and the test results for 170GHz gyrotron in KSTAR

ECH system.

Experimental characterization of quasilinear eects on ECRH

J. Decker1, S. Coda2, T. Goodman2, J. Kamleitner2, Y. Peysson1, Y.-S. Bae3, J. Jeong3, M. Joung3

1CEA, IRFM, F-13108, Saint-Paul-lez-Durance, France.2EPFL, CRPP, Association EURATOM - Confédération Suisse, CH - 1015 Lausanne, Switzerland.

3National Fusion Research Institute, Daejeon 305-333, Korea

Resonant interaction between electrons and radio-frequency (RF) waves can lead to a signicant distortionof the electron distribution function, which in turn aects the damping rate and current drive eciency of theRF waves. This eect cannot be described by linear modelling of electron cyclotron resonant heating (ECRH)and current drive (ECCD), but requires instead to solve the Fokker-Planck (FP) equation with a quasilinear(QL) treatment of the wave-particle interaction as provided for example by the code LUKE [1] associatedwith the ray-tracing code C3PO [2]. C3PO/LUKE simulations predict that the so-called quasilinear eectson EC damping result from a competition between the enhancing eect of suprathermal electrons, and thereducing eect due to attening of the electron distribution in momentum space. Besides, the current drivenby EC waves is also a balance between the competing Fisch-Boozer [3] and Ohkawa [4, 5] eects describing,respectively, the asymmetric resistivity and the magnetic trapping resulting from perpendicular momentuminjection.

Quasilinear eects have been invoked to explain the abnormally high optical depth of extraordinary ECwaves at third harmonic (X3) measured in TCV in the presence of X2 heating and CD [6]. However, FPQLmodelling of these experiments was not conclusive. The present work reports on dedicated experimentsdesigned specically to characterize quasilinear eects predicted by FPQL modelling. In the TCV tokamak,the EC optical depth is determined either from measurements of the stray emission [7] or derived from thestored energy via the diamagnetic loop (DML) in the presence of power modulation [6]. In addition, a newtomographic system of hard X-ray detectors is used to measure fast electron bremsstrahlung and characterizethe population of suprathermal electrons [8]. With plasma parameters adjusted so that the absorption ofthe ordinary (O) mode at second harmonic is incomplete, enhanced O2 absorption is demonstrated in thepresence of suprathermal electrons generated by X2 ECCD.

The KSTAR tokamak oers an optimal conguration to demonstrate Ohkawa current drive as the X2power deposition of 170 GHz ECRH with a magnetic eld on axis up to 3.5 T is ideally located at mid-radiuson the low-eld side. Variations in the loop voltage are used to identify and characterize current drive eects.As predicted, it is found that the Ohkawa eect dominates over the Fisch-Boozer eect in this congurationsuch that a current is driven in the direction opposite to that of conventional ECCD.

References

[1] J. Decker and Y. Peysson, Euratom-CEA report EUR-CEA-FC-1736, 2004

[2] Y. Peysson, et al., Plasma Phys. Control. Fusion 53 124028 (2011)

[3] N. J. Fisch and A.H. Boozer, Phys. Rev. Lett. 45 720 (1980)

[4] T. Ohkawa, Nucl. Fusion 10 185 (1970)

[5] J. Decker, AIP Conference proceedings 694 447 (2003)

[6] S. Alberti, et al, Nucl. Fusion, 42 42 (2002)

[7] T. Goodman, et al., Development of Electron Cyclotron Wave Absorption Measurement for Real-Time

Polarization Optimization and Studies of Quasilinear Eects, 24th IAEA Fusion Energy Conference, SanDiego, USA 8-13 October 2012

[8] J. Kamleitner, et al., Study of suprathermal electron dynamics by energy-resolved tomography of hard

X-ray emission on the TCV tokamak, 40th European Physical Society Conference on Plasma Physicsand Controlled Fusion, Espoo, Finland, 1-5 July 2013, p. P5.127

1

Study of synergetic effect of X2 and X3 EC Wave in KSTAR

Y. S. Baea, J. Decker

b, J. H. Jeong

a, K. D. Lee

a,

aNational Fusion Research Institute, Daejeon 305-333, Korea

bAssociation Euratom-CEA, DSM/IRFM, CEA Cadarache, 13108 St Paul lez Durance,

France

e-mail address: ysbae@nfri.re.kr

The electron cyclotron (EC) wave at the third harmonic extraordinary (X3) mode is optically

thin with the lower absorption rate than the EC wave at the second harmonic extraordinary

(X2) mode by a factor of kbTe/mc2. Therefore, the high electron temperature plasma is

favorable to have better X3 EC wave absorption. But TcV experiment showed synergetic

effect of X3 EC wave absorption when X2 electron cyclotron current drive (ECCD) is applied

together [1]. According to TcV experiments, there is evidence that some of the injected X3 EC

power is absorbed by suprathermal electrons in an energetic tail created by the X2 ECCD

preheating.

An experimental study of the X-mode absorption at the second and the third harmonic

frequencies has been performed in KSTAR tokamak. The X2 EC frequency is 110 GHz and

the X3 EC frequency is 170 GHz at the nominal KSTAR operating toroidal magnetic field.

From the 1-D model of the synergetic effect, the X3 cold resonance should lie at low field

side with the X2 cold resonance at the high field side to meet the condition which both X2

and X3 EC waves interact with the same resonant electron at the same radial position.

However, 170 GHz X3 cold resonance lies at the high field side with distance of 54 mm from

the X2 cold resonance position in KSTAR. But, the more recent experiment and modeling

study in TcV shows that there is another way of X3 absorption enhancement by X2 EC

heating through the same flux surface [2]. This paper presents the study of the synergetic

effect of X3 absorption by the X2 ECCD with a scheme of two beam target positions at the

same flux surface by scanning the poloidal and toroidal beam injection angles to enhance X3

absorption even in the low temperature plasma in KSTAR. For this study, the 3D relativistic

ray/tracing and Fokker-Planck codes C3PO/LUKE is used for the quantitative prediction of its

synergetic effect. The C3PO/LUKE codes are appropriate for X2 and X3 synergy calculations

as the distribution function can be calculated with either one or both waves affecting the

absorption rate of each other. This paper also introduces the future ECH system upgrade plan.

[1] S. Alberti et al., Nucl. Fusion 42 (2002) 42-45.

[2] S. Gnesin et al., Plasma Phys. Control. Fusion 54 (2012) 035002.

Demonstration of sawtooth period control with EC waves in KSTAR plasma

J. H. Jeonga, Y. S. Baea, M. Jounga, D. Kimb, T. P. Goodmanb, O Sauterb, K. Sakamotoc,

K. Kajiwarac, Y. Odac, J. G. Kwaka, W. Namkungd, M. H. Chod and H. Parke

a National Fusion Research Institute, Daejeon, Korea b CRPP - EPFL, Lausanne, Switzerland,

c Japan Atomic Energy Agency, Naka, Japan d Department of Physics, POSTECH, Pohang, Korea

e School of Electrical and Computer Engineering, UNIST, Ulsan, Korea

Email address: jhjeong@nfri.re.kr

The sawtooth period control in tokamak is important issues in recent years because the sawtooth crash can trigger TM/NTM instabilities and drive plasmas unstable. The control of sawtooth period by the modification of local current profile near the q=1 surface using ECCD has been demonstrated in a number of tokamaks including KSTAR. As results, developing techniques to control the sawtooth period as a way of controlling the onset of NTM has been an important area of research in recent years. In 2012 KSTAR plasma campaign, the sawtooth period control is carried out by the different deposition position of EC waves across the q=1 surface. The sawtooth period is shortened by on-axis co-ECCD (destabilization), and the stabilization of the sawtooth is also observed by off-axis co-ECCD at and outside q=1 surface. In 2013 KSTAR plasma campaign, the sawtooth locking experiment with periodic forcing of 170 GHz EC wave is carried out to control the sawtooth period. The optimal target position which lengthens the sawtooth period is investigated by performing a scan of EC beam deposition position nearby q=1 surface at the toroidal magnetic field of 2.9 T. The injection-locking by the modulated EC beam is successfully demonstrated with the scan of modulation-frequency and duty-ratio at the low beta (βN~0.5) plasma. In this paper, the sawteeth behavior by the location of EC beam and the preliminary result of the sawtooth locking experiments in KSTAR will be presented. In addition, upgrade plan of EC antenna system including the state-of-art KSTAR 170 GHz ECH system will be presented.

Feedback-controlled NTM stabilization on ASDEX Upgrade

J. Stober, L. Barrera Orte, K. Behler, A. Bock, A. Buhler, H. Eixenberger, L. Giannone,

M. Maraschek, A. Mlynek, F. Monaco, E. Poli, Ch. Rapson, M. Reich, M. Schubert,

W. Treutterer, D. Wagner, H. Zohm and the ASDEX Upgrade Team

Max-Planck-Institut für Plasmaphysik, Garching, Germany

W. Kasparek

IGVP, Universität Stuttgart, Stuttgart, Germany

On ASDEX Upgrade feedback-control of neoclassical tearing modes (NTMs) using fast movable

ECRH launchers has been demonstrated. Real-time correlation analysis of 1 MHz ECE- and

Mirnov-data allows to determine the mode position in real-time in ECE-frequency space. Real-

time equilibrium reconstruction is used 3-fold: it allows to convert ECE-frequencies to flux

surface labels, it is used to reconstruct the density profile from interferometer data and together

with the latter it is needed to calculate the location of the ECCD deposition in space using ray-

tracing. Moving the ECCD location to the flux surface where the island is located makes even

fully developed islands disappear within a few τE. Though the concept may sound simple, major

efforts were necessary to get the total cycle time down to a few ten milliseconds, including the

concept of generalized real-time diagnostics which extract the relevant quantities (mode positions,

etc.) from massive data streams and pass them through a shared memory to the discharge control

system. For the ray tracing the code had to be parallelized to obtain the location of the ECCD for

four beams and its derivative with respect to the launcher angle within the required cycle time.

This contribution describes the successful implementation of the NTM stabilization concept as

sketched above and the remaining difficulties, mainly related to the necessary accuracy. This may

be overcome by a fine tuning of the launcher angles based on the mode amplitude. Further

discussed is the beneficial effect of modulated ECCD, i.e. power modulation in phase with the

island. Previous results obtained with phase locked ECCD modulation using feed forward Bt-scans

to vary the ECCD deposition could be qualitatively reproduced. Since in our actual set-up the

radial width of the ECCD-profile is typically smaller than the saturated island width, the effects of

modulation are rather small. Still O-point heating with 50% duty cycle is more efficient than CW-

heating (100% duty cycle), which in turn is more efficient than X-point heating with 50% duty

cycle. Experiments using a fast directional switch to modulate the power between two launchers

which target the mode at locations at which its phase varies by π are described in [1].

[1] W. Kasparek et al., “FADIS”, abstract submitted for selection for this Workshop

BEAM PROPAGATION AND STRAY RADIATION IN THE ITER EC H&CD UPPER LAUNCHER

P. Platania1, A. Bruschi1, D. Farina1, L. Figini1, T. P. Goodman2, A. Krause2 M. A. Henderson3, A. Moro1, G. Saibene4, C. Sozzi1, M. Toussaint2

1Istituto di Fisica del Plasma CNR, Euratom-ENEA Association, 20125 Milano, Italy

2Centre de Recherche en Physique des Plasmas, CRPP-EPFL, CH-1015 Lausanne, Switzerland 3ITER Organization, 13108 Saint-Paul-les-Durance, France

4Fusion for Energy, Barcelona, Spain

The four ITER Electron Cyclotron (EC) Upper Launchers (UL) are designed to control Magneto-Hydrodynamic (MHD) instabilities with the deposition of Electron Cyclotron power in the affected region of the plasma. According to the present design, each launcher comprises two rows of four input waveguides, whose output beam is focused and driven towards the plasma by four sets of mirrors. In order to study beam propagation, detrimental effects, straylight behaviour and to verify analytical calculations, a 3D model of the UL optical system has been implemented with the electromagnetic code GRASP®. Calculations are performed with the Physical Optics method: a known electromagnetic field propagates from a source to the next scattering surface inducing currents on the reflector that irradiates a new field; the procedure is repeated up to the last scatterer. Detailed description of the components are introduced: pure hybrid mode HE11 from cylindrical waveguide as input beams, real shapes of the mirror contours, semi-analytical description of the ellipsoidal surfaces of focussing mirrors. A conceptual calculation scheme for the entire UL system has been developed in order to take into account not only the direct contribution of the single source on its next scatterer but also the first order indirect effects: crosstalk from different lines of the same row and crosstalk from different rows. The distortions due to crosstalk have been evaluated after reflection on the first (M1) and third (M3) set of mirrors. The mirror modelization could have in principle an impact on the results of the electromagnetic calculations; the consequence of reflection on single line (one per beam) and multisurface (from merged mirrors, one per row) scatterers has been considered and has to be further studied. Regarding the stray radiation, the main directions of propagation from and around the steering mirrors have been identified and the impact on surrounding structures quantified. The model allows us to evaluate the level of stray radiation due to the crosstalk effect among different possible parasitic optical paths. Evaluations on beam shapes and the degree of superposition at targets are compared with results from gaussian beam propagation analytical calculations. The evaluations presented here have been performed on the preliminary UL design, the last major milestone before finalization; however, the numerical model is suitable to be applied to future evolutions of the setup and/or other configurations. This work was supported by Fusion for Energy under the grant contract No. F4E-2010-GRT-161. The views and opinions expressed herein reflect only the author’s views. Fusion for Energy is not liable for any use that may be made of the information contained therein.

Instrumentation & control system architecture for ITER Electron cyclotron

heating and current drive plant system Dharmesh Purohita , Carlo Sozzie, Caroline Darbosa, Dhananjay Billavad, Franco Gandinia, Filippo Sartorib , Gabriella Sabienneb,

Gustavo Granuccie, Mario Cavinatob Mark Hendersona, Thibault Gassmana, Toshimichi Omoria, Yasuhisa Odac

a. ITER Organization, St. Paul-lez-Durance, 13115 France;

b. Fusion for Energy, C/ Josep Pla 2, Torres Diagonal Litoral-B3,E-08019 Barcelona – Spain

c. Japan Atomic Energy Agency (JAEA) 801-1 Mukoyama, Naka-shi, Ibaraki 311-0193 Japan

d. Tata Consultancy Services La Defense 8, 2800 Puteaux France;

e. Istituto di Fisica del Plasma CNR, Euratom-ENEA Association, 20125 Milano, Italy

The ITER Electron Cyclotron Heating and Current Drive system (ECH &CD) consists of 24 Radio

Frequency sources (Gyrotrons), 12 sets of high voltage Power supplies, and 24 transmission lines (TL)

which connect Gyrotron to either of 4 upper launchers or 1 equatorial launcher. The ECH &CD system is

procured from 5 domestic agencies (DA) contributing via more than 11 in-kind procurement

arrangements. All the subsystems being provided by various DAs will have their own control unit, while

the plant/main controller of ECH &CD system (ECCS) will be procured from Europe domestic agency

(EU-DA). The deliveries of various subsystems and installation are spread out across a long duration,

which dictates a long duration of deliveries for the control system as well.

The control system design is being developed to incorporate the staged commissioning and lifecycle

requirement of ECH &CD system to ITER machine. A functional approach has been used to define

functions & responsibility sharing of each subsystem control unit and EC plant controller (ECPC). The

control system architecture and interfaces are defined to best optimize the procurement scope, staged

installation and commissioning of various subsystems, integrated operation of ECH &CD system for

ITER machine and also allowing future upgrade of system to incorporate advance control functions. The

ECH &CD instrumentation and control system (ECCS) is split into various hierarchy of control viz RF

Generation, Plant configuration management, RF Launching and investment protections. The HVPS are

given highest autonomy to test and operate as they are first to arrive at ITER and will be installed;

however the embedded functionality internal to the power supply will be enable it to be integrated as

slave to Gyrotron control unit. HVPS control will be done dynamically by the Gyrotron controller as well

as ECPC depending on operation needs. The architecture developed and interfaces defined between

HVPS, Gyrotron and ECPC will ensure efficient operation during the diverse scenarios, while

maintaining high reliability and modularity. ECPC will also be communicating with Plasma control

system of ITER for defining as well as controlling the Gyrotron operation. Gyrotron control unit will be

supplied by 4 domestic agencies, while HVPS set will be delivered from 3 contributors. EC plant

controller will also be responsible to manage the configuration of plant and optimize the power launched

from launchers. The launcher control and high level control of Gyrotron are considered as functions for

experiments, which necessitates a highly flexible architecture. The control system allows this flexibility

as part of plant controller dissociating with engineering control, which are defined as standard practices

of operating HVDC, Gyrotron, Switches, Steering mirror actuator and protection of the plant components

based on status received from subsystem. The architecture is configured to reduce traffic of signals

between the subsystem control units, as well as commands by aggregating the signals at a local level. The

protection function will also be mapped according to operation scenario and configuration ECH &CD

plant. More than 15 operation scenarios have been defined to incorporate all the needs of ECH &CD

system in the various configurations and limit the manual operation.

The resulting functional architecture of ECH &CD control system is detailed in this paper. It would

be worth to note that ITER-CODAC group has developed the standards of software and hardware, which

have been used to develop the architecture; however the architecture of ECCS has also introduced

industrial solutions to meet the demanding requirement of ECH &CD. A test and verification program

under the prototype activity is being performed at Japan Atomic Energy Agency (JAEA) gyrotron test

stand to validate the interfaces and architecture, and later on update accordingly. The paper highlights

functional study, architecture design of EC control system, CODAC standards, and challenges to ECCS

and mitigation strategy.

Development of Real-Time Neoclassical Tearing Mode Control System with ECH/ECCD in KSTAR

M. Joung, M. H. Woo, J. H. Jeong, S. H. Hahn, S. W. Yun a, W. R. Lee, J. G. Bak, K. D. Lee, Y. S. Bae, J. G. Kwak, W. Namkung b, H. Park c, J. Hosea d, R. Ellis d, K. Sakamoto e, M. H.

Kim f, Y. S. Na f, M. J. Choi g, G. S. Yun g, A. S. Welander h, D. A. Humphreys h and J. Lohr h

National Fusion Research Institute, Daejeon, Korea a ITER Organization, Cadarache Center, France

b Pohang Accelerator Laboratory, Pohang, Korea c School of Electrical and Computer Engineering, UNIST, Ulsan, Korea

d Princeton Plasma Physics Laboratory, Princeton, USA e Japan Atomic Energy Agency, Naka, Japan

f Department of Nuclear Engineering, SNU, Seoul, Korea g Department of Physics, POSTECH, Pohang, Korea

h General Atomics, San Diego, USA

E-mail address: whitemi@nfri.re.kr

A real time Neo-classical Tearing Mode (NTM) control system using ECH/ECCD is under development to achieve a high performance and steady state plasma in KSTAR. Since the NTMs are inevitable in high beta plasma, it degrades the plasma performance and finally causes a serious disruption. In KSTAR, it is experimentally observed that the NTM is triggered by a big sawtooth crash in the low βN (about 0.8) plasma, and then the plasma beta is decreased. Since the auxiliary heating power increases in KSTAR, the NTMs will appear more frequently in the future. The NTMs can be effectively affected by the localized heating and current drive to produce the bootstrap current inside the NTM island. For the NTM suppression, the ECH/ECCD is one of the powerful and well-known tools because the power deposition can be highly localized and controllable by steerable antenna. KSTAR has a plan to develop a real time NTM control system by using ECH/ECCD. It will be implemented in the basis on KSTAR Plasma Control System (PCS) including active feedback control of ECH system. This work is mainly for a good alignment the ECH/ECCD deposition position with the NTM island location. Active feedback control of ECH includes feedback control of the deposition position of ECH power by the fast mirror movement, real-time ray-tracing calculation of EC beam, and real-time detection of the deposition location of EC power. The NTM island location tracing will be fulfilled by real-time detection of mode, position, and phase of NTMs by real-time equilibrium reconstruction data and diagnostics such as Mirnov coils, ECE, MSE, and so on. In 2013, we had set up real-time ECH mirror movement during a single plasma discharge by upgrading the KSTAR PCS and the ECH launcher controller. The KSTAR 170 GHz ECH launcher, designed and fabricated in collaboration with PPPL and POSTECH, has a final steerable mirror in both poloidal and toroidal direction. Only poloidal direction of the steerable mirror was controlled by the PCS and then ECH power deposition position was changed along the resonance layer. The slow PLC launcher controller has been upgraded to FPGA-based system for reducing the response time. The test results of the launcher controller speed will be presented. The first NTM suppression experiment was performed with the PLC launcher controller with long response time. In this paper, we present experimental results of NTM suppression, NTM characteristics in present KSTAR, and upgrade plan of ECH and the real time NTM control system.

Direct measurement of power and refracted trajectory of transmitting electron

cyclotron beam through plasma on the Large Helical Device

H. Takahashi, S. Kubo, T. Shimozuma, H. Igami, Y. Yoshimura, M. Nishiura, S. Ito, S. Kobayasi,

Y. Mizuno, K. Okada, S. Kamio, R. Makinoa, S. Ogasawara

a, T. Mutoh, M. Osakabe, K. Nagasaki

b

and the LHD experiment group

National Institute for Fusion Science, Toki, Japan

aDepartment of Energy Engineering and Science, Nagoya Univ., Nagoya, Japan

bInstitute of Advanced Energy, Kyoto Univ., Uji, Japan

takahashi.hiromi@LHD.nifs.ac.jp

In toroidal plasmas, it is important to evaluate the single path absorption power for efficient electron

cyclotron resonance heating (ECRH). In order to measure the transmission power of EC beam,

which is not absorbed into the plasma in the single path, a target plate facing on the ECRH antenna

has been set in the vacuum vessel on the Large Helical Device (LHD).

The target plate ( = 364 mm, t = 5 mm) is made of isotropic graphite. The plate has been set 1.7 m

lower from the equatorial plane of the LHD and receives the EC beam ejected from the facing-on

upper launcher. The temperature rise of the plate due to the EC beam has been measured using an

IR camera. This system has the advantage of the abilities to evaluate the transmitting EC beam

power and to measure the refraction of the EC beam in the magnetically confined plasma.

In the experiments, 77-GHz EC beam was injected to the plasma as the O mode under the magnetic

field of 1.375 T, where the EC beam absorption to the plasma was expected to be considerably

small. Clear dependence of the beam refraction on the electron density gradient was observed from

the landing-point of the EC beam on the target plate. The displacement of the EC beam trajectory

due to the refraction well agreed with the ray trace calculation in the toroidal direction but not in the

radial direction. The effect of the dielectric tensor and the peripheral density profile on the ray trace

calculation will be discussed in the presentation.

Long-pulse plasma discharges by upgraded ECRH system in LHD

Y. Yoshimura, H. Kasahara, K. Nagasaki1, M. Tokitani, N. Ashikawa, Y. Uede2, S. Ito, S. Kubo,

T. Shimozuma, H. Igami, H. Takahashi, M. Nishiura3, S. Kobayashi, Y. Mizuno, K. Okada,

S. Ogasawara, R. Makino, I. Yamada, K. Tanaka, T. Mutoh, H. Yamada

National Institute for Fusion Science, Toki, Japan

1) Institute of Advanced Energy, Kyoto University, Kyoto, Japan

2) Graduate School of Engineering, Osaka University, Osaka, Japan

3) Graduate School of Frontier Sciences, The University of Tokyo, Kashiwa, Japan

First author email address: yoshimura.yasuo@LHD.nifs.ac.jp

Until 2009, three high-power, over 1 MW each, 77 GHz gyrotrons have been installed and applied

to LHD experiment [1, 2]. In addition, a 154 GHz gyrotron of 1 MW was installed in 2012.

The 77 GHz gyrotrons suffer gradual increases of internal pressure during long-pulse operation

delivering power to LHD. To mitigate the problem, quasi-steady operation by combination of on-

off operations of the 77 GHz gyrotrons was performed. Using two 77 GHz gyrotrons alternately at

intervals of two minutes and an 84 GHz gyrotron continuously, a 30 min. long-pulse discharge with

the line average electron density ne_ave of 0.7×1019 m−3 and the central electron temperature Te0 of

1.5 keV was achieved by the time average injection power Pinj of 240 kW in 2012, showing

significant progress in sustained density from the former 65 min. discharge with ne_ave of 0.15×1019

m−3 and Te0 of 1.7 keV by Pinj of 110 kW of 84 GHz wave [3].

In 2013, one of the 77 GHz gyrotron was improved to furnish a sub-window to remove stray

radiation inside the tube. And the new 154 GHz gyrotron was applied to long-pulse discharge

experiment. Using 3 gyrotrons: 154, the improved 77 and 84 GHz with Pinj of 380 kW in total,

higher temperature plasma having steep temperature gradient typical for internal transport barrier

with ne_ave of 0.8×1019 m−3 and Te0 of 4 keV was stably sustained for 12 min.

References

[1] H. Takahashi et al., 2010 Fusion Sci. Technol. 57 19

[2] T. Shimozuma et al., 2010 Fusion Sci. Technol. 58 530

[3] Y. Yoshimura et al., 2010 Fusion Sci. Technol. 58 551

Efficient ECRH mode excitation through inhomogeneous peripheral plasmain LHD

S. Kubo1,2, H. Igami1, T. Shimozuma1, H. Takahashi1, Y. Yoshimura1, M. Nishiura3,R. Makino2, S. Ogasawara2, T. Mutoh1 and LHD experiment group

1National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan2Dept. Energy Engineering and Science, Nagoya University, Chikusa 464-8603, Japan3Grad. School of Frontier Sciences, The University of Tokyo, Chiba 277-8561, Japan

kubo@LHD.nifs.ac.jp

In order to realize an efficient ECRH and also to reduce stray radiation due to non-absorbed powerduring ECRH, it is necessary to excite a wave that is absorbed well near the electron cyclotronresonance. In the normal fusion magnetic field confinement machine and in the electron cyclotronfrequency range, WKB approximation is valid almost all the way from antenna to the absorptionregion due to the large scale length of the plasma density λn, magnetic field strength λB and themagnetic shear λs as compared with the local wavelength λ[1]. In these situation, it is well knownthat the O/X mode propagates as O/X mode if λs ≫ λ. Even in these situation, if λs and λn

are comparable, there still remains the question from where ”X”- or ”O”- mode become ”X”- or”O” mode at the peripheral region [2]. This situation become more critical in the case of obliqueinjection through stochastic magnetic field region in LHD where λs ∼ λn.In order to simulate this situation, one dimensional full wave calculation code which solve elec-tromagnetic wave equation under arbitrary magnetic field configuration and arbitrary density profilefor a given polarization state are developed and incorporated in the ray tracing code ”LHDGauss”[3].”LHDGauss” is now upgraded to include experimental Thomson scattering profile data and threedimensional equilibrium mapping[4] and extrapolating equilibrium and density profile to the stochas-tic magnetic field region.Recent experimental results especially for oblique injection in the relatively higher density regimeare compared with those from calculated one and the validation of the method will be discussed.

[1] I. Fidone and G. Granata, Nucl. Fusion 11 (1971) 133.

[2] T. Notake, S. Kubo et al., Plasma Phys. Control. Fusion 47 (2005) 531.

[3] S. Kubo, H. Idei, T. Shimozuma, Y. Yoshimura et al., in Proceedings of 11th Int. Congress onPlasma Physics (July 2002, Sydney, Australia) (2002) 133.

[4] C Suzuki, K Ida, Y Suzuki, et al. Plasma Phys. Control. Fusion 55 (2013) 014016.

Power measurement system of ECRH on HL-2A

He Wang, Jun Rao, Chao Wang, Gangyu Chen, Jun Zhou, Mei Huang, Zhihong Lu Southwestern Institute of Physics, Cheng Du, China

Abstract:

Electron Cyclotron Resonance Heating (ECRH) is one of the main heating methods for HL-2A tokamak. The ECRH system with total output power 5MW has been equipped on HL-2A which include 6 sets of 0.5MW/1.0s at a frequency of 68GHz and 2 sets of 1MW/3s at a frequency of 140GHz. The power is an important parameter for ECRH system during commissioning and ECRH experiment. In this paper, the method for measuring the power of ECRH system on HL-2A is introduced which include calorimetric techniques and directional coupler.

Usually, the high pulsed microwave power was measured by calorimetric techniques during commissioning of the gyrotrons and ECRH experiments for us. The principle and the process of calorimetric method are introduced in this paper. This power measuring system consists of calorimeter, electronic unit and timing power supply, and measured data are recorded and processed by a data acquisition and process system. The calorimeters were installed on Match Optical Unit (MOU), gyrotron output window and the window near the EC launcher separately in each ECRH system. Firstly, the percentage of spurious power absorbed by the cooling water in MOU is measured in commissioning of the gyrotron, which is about 7%~11% of total output power of gyrotron for 68GHz system and about 5% for 140GHz system. Then the output power of gyrotron could be calculated by this spurious power. The percentage of power absorbed by the window is also obtained in commissioning which is about 1.05% for BN window in 68GHz system. And the transmission efficiency of ECRH system is about 90% by measuring the absorbed microwave power in the MOU, gyrotron output window and the window near the EC launcher of the 68GHz system.

Now we are designing directional coupler which bases on the theory of electromagnetic coupling through apertures, and it is a new method for us. The directional couplers can be installed in the miter bend and measure the power in real time. They have been machined in the factory, and we hope the directional couplers can be test in the next experiment period. Key words: HL-2A, ECRH, Power measurement

A model of multi-pass absorption of external EC radiation

at initial stage of discharge in ITER

P.V. Minashin1, A.B. Kukushkin

1, R.R. Khayrutdinov

1,2, V.E. Lukash

1

1Tokamak Physics Institute, NRC Kurchatov Institute, Moscow, Russia

2Troitsk Institute for Innovation & Fusion Research, Troitsk, Russia

Due to technological issues the ohmic plasma breakdown in tokamak-reactor ITER is only

possible over a narrow range of plasma pressure and magnetic field errors [1]. In this

connection for the reliable plasma start-up in ITER it is planned to use electron cyclotron

resonance heating (ECH, ECH-assisted start-up) [2], [3]. The ECH is a standard way for

plasma start-up in stellarators and already showed to be an effective method for plasma

breakdown in tokamaks [4].

Modelling of the initial stage of plasma discharge in ITER with the 0D model [1] showed

that in a wide range of initial conditions, taking into account beryllium impurities, the 3 MW

of absorbed external EC radiation is needed to achieve the plasma breakdown (for the carbon

impurities even 5 MW of absorbed power is not enough). However, in [1] the efficiency of

EC absorption was not calculated. Recent 1D simulations [5] of the ECH start-up in ITER

with the help of the OGRAY code [6] for the ECH calculations showed that the planned EC

power could be not enough for plasma breakdown due to the low efficiency of the single-pass

ECH power absorption.

Here we propose a model for calculating the efficiency of the absorption of external EC

power in tokamaks at initial stage of discharge. The single-pass absorption of injected EC

radiation is evaluated with the scaling [5] obtained using the OGRAY code. For the

subsequent multi-pass absorption, after first reflection of the EC wave from the wall of the

vacuum chamber, we develop a model based on the assumption of the isotropy/uniformity of

the respective EC radiation intensity in plasma (a semi-analytical solution of the radiative

transfer problem for the case of multiple reflection of radiation from the wall). The model

modifies the approach of the CYNEQ code [7], [8] developed for the plasma-produced EC

radiation transport at high EC harmonics and verified in the benchmarking [9], [8], [10].

In the frame of this approach, we consider the following case: (a) multiple reflection of

injected EC wave (O-mode) from the wall; (b) polarization scrambling in wall reflections;

(c) full single-pass absorption of the X-mode. Our parametric analysis of the efficiency of

multi-pass absorption of injected EC radiation for typical values of the electron temperature

and density at the initial stage of discharge in ITER shows strong dependence on the O-X

conversion in wall reflections.

References

1. B. Lloyd, P.G. Carolan, et al., Plasma Phys. Controlled Fusion 38, 1627-1643 (1996)

2. ITER Physics Expert Group on Energetic Particles, Heating and Current Drive, ITER Physics

Basis Editors, Nuclear Fusion 39, 2495 (1999)

3. T. Omori, M.A. Henderson, et al., Fusion Engineering and Design 86, 951-954 (2011)

4. J. Stober, G.L. Jackson, E. Ascasibar, Y.S. Bae, et al., Nuclear Fusion 51, 083031 (2011)

5. R.R. Khayrutdinov, A.Y. Kuyanov, V.E. Lukash, et al., 38th EPS Plasma Conf. P2.085 (2011)

6. A.V. Zvonkov, A.Y. Kuyanov, A.A. Skovoroda, A.V. Timofeev, Plasma Physics Reports 5, 389-

400 (1998)

7. A.B. Kukushkin, 14th IAEA Conf. Plasma Physics 2, 35 (1992)

8. A.B. Kukushkin , P.V. Minashin, 24th IAEA Fusion Energy Conf. TH/P6-25 (2012)

9. F. Albajar, M. Bornatici, F. Engelmann , A.B. Kukushkin, Fus. Sci&Tech. 55, 76 (2009)

10. A.B. Kukushkin, P.V. Minashin , A.R. Polevoi, Plasma Phys. Reports 38, 211 (2012)

Engineering aspects of design and integration of microwave diagnostics in ITER

V.S. Udintsev1, G. Vayakis1, M.J. Walsh1, J.-M. Drevon1, T. Giacomin1, J.W. Oosterbeek2, M. Portalès1, A. Sirinelli1, M.E. Austin3, R. Feder4, M.A. Henderson1, D. Johnson4, H. Pandya5, P. Sanchez6, D. Shelukhin7, G. Taylor4, V. Vershkov7 1ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance, France 2Technical University of Eindhoven, The Netherlands 3Fusion Research Center, University of Texas at Austin, Austin, TX, USA 4Princeton Plasma Physics Laboratory, USA 5Institute for Plasma Research, Bhat, Gandhinagar-382 428, India 6Fusion for Energy, ITER Department, Diagnostic group, Barcelona, Spain 7NFI, RRC Kurchatov Institute, Moscow, Russian Federation ITER ECE diagnostic [1] needs not only to meet measurement requirements, but also to withstand various loads, such as electromagnetic, mechanical, neutronic and thermal, and to be protected from stray ECH radiation and other millimeter wave emission, like Collective Thomson scattering which is planned to operate at 60 GHz. Same or similar loads will be applied to other millimetre-wave diagnostics [2], located both in-vessel and in-port plugs. These loads must be taken into account throughout the design phases of the ECE and other microwave diagnostics to ensure their structural integrity and maintainability. The integration of microwave diagnostics with other ITER systems is another challenging activity which is currently ongoing through port integration and in-vessel integration work. Port Integration has to address the maintenance and the safety aspects of diagnostics, too. Engineering solutions which are being developed to support and to operate ITER microwave diagnostics, whilst complying with safety and maintenance requirements, will be addressed and discussed in the paper and during a dedicated ITER ECE session scheduled in the Workshop’s agenda. [1] G. Taylor et al., this Workshop (2014). [2] V.S. Udintsev et al., in Proc. of EC-17 Workshop, Deurne, The Netherlands (2012).

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating

O-1 ECE Imaging of Edge-Localized Modes in the High-Field Side†

G. S. Yun1*

, J. Lee1, G. Choe

1, M. J. Choi

1, W. Lee

1,

H. K. Park2, S. H. Hahn

3, K. D. Lee

3, S. W. Yoon

3, N.C. Luhmann Jr.

4

1Pohang University of Science and Technology, Pohang, Gyeongbuk 790-784, Korea

2Ulsan National Institute of Science and Technology, Ulsan 689-798, Korea

3National Fusion Research Institute, Daejeon 305-333, Korea

4University of California, Davis, CA 95616, U.S.A

*E-mail: gunsu@postech.ac.kr

Edge-localized modes in the high-field side (HFS) of H-mode plasmas have been visualized in

2-D for the first time on the KSTAR tokamak by measuring the fundamental ordinary mode

(O-1) electron cyclotron emissions (ECE). A numerical analysis on the emission and absorption

profiles in the HFS pedestal region, where the optical depth is rapidly varying, was performed to

check the validity of local measurement which is necessary for imaging. The analysis showed

that the coherent density perturbations in the HFS pedestal region can be imaged by O-1 ECE.

The apparent amplitude and poloidal spacing of the ELM filaments are comparable in the HFS

and LFS, which may not be accountable by any existing numerical simulation of ELMs. Other

peculiarities found in the HFS ELM images are analyzed such as weaker amplitude at the

mid-plane and the apparent filament position being outside of the separatrix. The latter may be

explained by a large paramagnetic increase of the toroidal field up to a few % and several

candidate mechanisms for the paramagnetism are proposed. A possible mixing of HFS O-1 ECE

with electron Bernstein wave emissions originating from the LFS is also discussed. †Work

supported by NRF Korea under grant no. NRF-2009-0082507 and US DoE under contract no.

DE-FG-02-99ER54531.

Correlation ECE diagnostic in Alcator C-Mod

C. Sung1, A. E. White

1, N. T. Howard

1, D. Mikkelsen

2, J. Irby

1, R. Leccacorvi

1, R. Vieira

1, C. Oi

1, J.

Rice1, M. Reinke

1, C. Gao

1, P. Ennever

1, M. Porkolab

1, R. Churchill

1, C. Theiler

1, J. Walk

1, J. Hughes

1,

A. Hubbard1, M. Greenwald

1 and the Alcator C-Mod team

1Plasma Science and Fusion Center, Massachusetts Institute of Technology, Cambridge,

Massachusetts, USA

2Princeton Plasma Physics Laboratory, Princeton, New Jersey, USA

Although ECE radiometers are widely used diagnostics for Te profile measurements, the radiometer

has a limitation when it comes to measuring turbulent Te fluctuations, due to inherent thermal noise in

the radiometer. Correlation ECE (CECE) is a diagnostic technique which allows measurement of

small amplitude Te fluctuations through standard cross-correlation analysis methods[1]. In C-Mod, a

new CECE diagnostic was designed and installed in 2011[2]. The initial CECE diagnostic was a 4ch

ECE radiometer measuring 2nd

harmonic X-mode EC emission (230-248GHz), which was upgraded

to an 8ch radiometer during the 2012 campaign. We have measured turbulent Te fluctuations in

various plasma conditions in auxiliary heated L-mode, I-mode and H-mode plasmas and ohmic

plasmas. In ohmic plasmas, we have observed for the first time that local Te fluctuations decrease

across the linear-to-saturated ohmic confinement transition, with fluctuations decreasing with

increasing plasmas density[3]. In I-mode plasmas, we observed the reduction of core Te fluctuations,

which indicates the change of turbulence occurs not only in the pedestal region but also in the core

across the L/I transition[4]. Gyrokinetic analysis is being performed to interpret these measurements,

and results from non-linear gyrokinetic simulations (GYRO) are being compared with CECE

measurements using a synthetic diagnostic.

References

[1] G. Cima et al, Physics of Plasmas, 2 720 (1995)

[2] C. Sung et al, Review of Scientific Instruments, 83:10E311 (2012)

[3] C. Sung et al, Nuclear Fusion, 53 083010 (2013)

[4] A. White et al, in 55th Annual meeting of APS-DPP, Abstract : CO4.00003 Vol 58 No 16 (2013)

Localised Microwave Bursts During ELM Events on MAST

S. Freethya, V. F. Shevchenko

a, B. K. Huang

b, R. G. L. Vann

c

a EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB, UK

b Centre for Advanced Instrumentation, Department of Physics, Durham University, Durham,

UK

c York Plasma Institute Department of physics, University of York, York, YO10 5DD, UK

Strong microwave bursts with an intensity more than a factor of 1000 greater than thermal

emission have been seen on MAST during ELM events. These events occur just as the Dα

signal begins to rise. Using the Synthetic Aperture Microwave Imaging (SAMI) array, high

time resolution phased array imaging reveals each ELM burst to be composed of many

shorter ~1µs events clustered together. The imaging shows these bursts to be highly spatially

localised and to originate preferentially from a few locations. It is thought that the origin of

these bursts may be electron heating during magnetic reconnection at filament detachment.

We will discuss possible mechanisms for the emission characteristics using the 2D +

frequency imaging as a strong constraint.

This work is supported by RCUK and EURATOM.

Spectroscopic diagnostics of superthermal electrons with high-number

harmonic EC radiation in tokamak reactor plasmas

P.V. Minashin, A.B. Kukushkin

Tokamak Physics Institute, NRC Kurchatov Institute, Moscow, Russia

DEMO, the next generation of tokamaks after ITER, will be characterized by substantially

higher plasma temperatures (more than 30 keV in plasma center), that may limit operational

spaces of the existing plasma diagnostics, including the electron cyclotron emission (ECE)

diagnostics [1]. Furthermore, strong auxiliary heating generates deviations of electron

velocity distribution (EVD) from a Maxwellian, distorting the basis of the ECE diagnostics of

temperature in tokamaks, namely the one-to-one correspondence between three quantities,

toroidal major radius coordinate, toroidal magnetic field, and low-number harmonic of EC

fundamental frequency of registered radiation [2]. Thus, for DEMO an extension of ECE

diagnostics is needed.

Here we suggest a method of spectroscopic diagnostics of superthermal EVD with the

high-number harmonic EC radiation. To calculate the spectrum of the outgoing EC radiation

we use the CYNEQ code [3], [4], which is based on a semi-analytical solution of the radiative

transfer equation for the case of multiple reflections of plasma-produced EC radiation from

the wall. The method relies on solving an inverse problem for reconstruction of the EVD in

parallel and perpendicular-to-magnetic field components of electron momentum at high and

moderate energies responsible for the emission of the high-number harmonic EC radiation.

The well-known difficulty of solving such a problem in the case of arbitrary EVD (radiative

transfer effects for low-number harmonics and nonlocality of transport for high-number

harmonics, see, e.g., [5] for diagnostics of superthermal electrons in tokamak T-10 at

relativistically down-shifted frequencies of 1st and 2

nd harmonics) restricted the capability to

reconstruct main parameters of the EVD to once most sensitive to EC radiation. We show the

possibility to restore the average transverse-to-magnetic field momentum of superthermal

electrons in central plasma, e.g., for Te(0)~30 keV we can restore the energy of ~2-3Te

superthermals from spectrum at 10-14 harmonics. The combination of such a diagnostic with

those which can determine the parameters of the bulk Maxwellian plasma (like Thomson

scattering diagnostics) may provide a tool for spectroscopic plasma diagnostics of fusion-

reactor plasmas.

References

1. A.J.H. Donné, A.E. Costley , A.W. Morris, Nuclear Fusion 52, 074015 (2012)

2. M. Bornatici, R. Cano, O.De Barbieri , F. Engelmann, Nuclear Fusion 23, 1153 (1983)

3. A.B. Kukushkin, 14th

IAEA Conference on Plasma Physics and Controlled Nuclear

Fusion Research. 2, 35 (1992)

4. A.B. Kukushkin , P.V. Minashin, 36th

EPS Conf. on Plasma Physics P-4.136 (2009)

5. P.V. Minashin, A.B. Kukushkin , V.I. Poznyak, EPJ Web of Conf. 32, 01015 (2012)

High Spatial Resolution ECE Measurements on DIII-D

M.E. Austin1 and D.D. Truong2

1University of Texas at Austin, 2100 San Jacinto Blvd, Austin, TX 78712-1047, USA 2University of Wisconsin-Madison, 1500 Engineering Dr., Madison, WI 53706, USA

A new set of channels has been added to the DIII-D ECE radiometer to make high spatial resolution Te measurements. Due to high optical depth in typical DIII-D plasmas, resolving sub-centimeter structure is possible. The high resolution channels are configured to use the same IF bands as the regular channels with eight fixed filters, having center frequencies separated by 250 MHz, in the 2–4 GHz range. Adjustable radial coverage is achieved by mixing higher IF frequencies down to the fixed filters’ frequency range. The channels are designed to be mostly useful in the low-field side edge region where modest Te values of 1–2 keV result in a minimum of relativistic broadening. For 200 MHz wide filters, emissivity calculations predict spatial widths of 6–8 mm for low-field-side edge region with optical depths exceeding 2 out to normalized radius r/a = 0.95. Expected uses of the channels will be to map out spatial dependence of Alfvénic eigenmodes, geodesic acoustic modes, and externally applied magnetic perturbations. Recent data obtained during NTM mode activity indicate the desired resolution is achieved.

This work was supported by the US Department of Energy under DE-FG03-97ER54415, DE-FG02-89ER53296, DE-FG02-08ER54999, and DE-FC02-04ER54698.

Development of Mode Conversion Waveguides at KITJianbo Jin, Gerd Gantenbein, John Jelonnek, Tomasz Rzesnicki, and Manfred Thumm

Karlsruhe Institute of Technology (KIT), Association EURATOM-KIT

Institute for Pulsed Power and Microwave Technology (IHM)

D-76131 Karlsruhe, Germany

As the highest-power coherent millimeter wave sources, gyrotrons are widely employed forElectron Cyclotron Resonance Heating (ECRH) and Current Drive (CD) in nuclear fusion plasmaexperiments. For high power gyrotrons operated in very high order cavity modes, internalQuasi-Optical (Q.O.) mode converters are used to transform the operating mode into a Gaussian mode.In general, a Q.O. mode converter consists of an oversized open-ended waveguide launcher and amirror system. In order to transform a high-order cavity mode into a Gaussian-like wave beam and toachieve low field at the edges of the launcher cuts, the launcher wall is dimpled to bunch the wavebeam.

Formerly, helically corrugated launchers have been developed at KIT [1]. For an operating mode,the azimuthal radiation width angle α can be defined as α=2arccos(Rc/R) where Rc and R are thecaustic and launcher radius respectively. In the case that the ratio b=2π/α is approximately an integer(e.g. 3), such a helically corrugated launcher can be designed to provide an RF beam with highGaussian mode content at the launcher aperture. As an example, Fig. 1 shows the wall profile designedfor a TE32,9-mode gyrotron, for which α≈124.4degree and b≈2.89. The results of mode matching revealthat the fundamental Gaussian mode content is about η =96.4% at the aperture.

Fig. 1 Wall profile (left) of the helically corrugated launcher and the field distribution (right).

A numerical method for the synthesis of mirror-line launchers has also been developed at KIT [2].A mirror-line launcher has been designed for a TE34,19-mode gyrotron. For the TE34,19 mode, the angleα≈142.4degree, then the ratio b=360degree/α≈2.53. The simulation results show the Gaussian modecontent of the RF beam is about 96.3% at the launcher aperture [2]. Hence, the limitation of the ratio bto be approximately an integer has been overcome in mirror-line launchers. A mirror-line launcher hasalso been designed for the ITER EU 1 MW TE32,9-mode gyrotron which is currently underdevelopment at KIT. Fig. 2 shows the wall profile and the field distribution. The fundamental Gaussianmode content of the RF beam is about 98.4% at the aperture.

Fig. 2 Wall profile (left) of the mirror-line launcher and the field distribution (right).

Recently, a new numerical method for launchers design has been developed [3]. The key point ofthe new method is to artificially generate a field distribution on the launcher wall as a target function.The wall profile of the launcher is iteratively optimized according to the field difference between theinput field and the target function. A launcher has also been designed in terms of the new method. Thewall profile and the field distribution are shown in Fig. 3. The fundamental Gaussian mode content ofthe RF beam is estimated as 97.7% at the aperture.

Fig. 3 Wall profile (left) of the launcher designed in terms of the new method and the fielddistribution (right).

This work, supported by the European Communities under the contract of Association betweenEURATOM and KIT, was carried out within the framework of the European Fusion DevelopmentAgreement. Part of this work is supported by Fusion for Energy (F4E) under grant contract No.F4E-GRT-432 and within the European Gyrotron Consortium (EGYC). EGYC is a collaboration ofCRPP, Swi tzerland; KIT, Germany; HELLAS, Greece; IFP-CNR, Italy. The views and opinionsexpressed herein do not necessarily reflect those of the European Commission or of F4E.

[1] J. Jin, M. Thumm, B. Piosczyk, T. Rzesnicki, “Theoretical investigation of an advanced launcher for a 2-MW 170-GHz

TE34,19 coaxial cavity gyrotron”, IEEE Transactions on Microwave Theory and Techniques, Vol. 54, No. 3, pp. 1139-1145,

2006.

[2] J. Jin., M. Thumm, B. Piosczyk, S. Kern, J. Flamm, and T. Rzesnicki, “Novel Numerical Method for the Analysis and

Synthesis of the Fields in Highly Oversized Waveguide Mode Converters,” IEEE Transactions on Microwave Theory and

Techniques, Vol. 57, No. 7, pp. 1661-1668, 2009.

[3] J. Jin, G. Gantenbein, J. Jelonnek, T. Rzesnicki, and M. Thumm, “A new method for the design of mode conversion

waveguides”, submitted for publication to IEEE Transactions on Microwave Theory and Techniques.

Remote-Steering Launchers

for the ECRH system on the Stellarator W7-X

W. Kaspareka, C. Lechte

a, B. Plaum

a, A. Zeitler

a,

V. Erckmannb, H. Laquab, M. Weißgerberb, A. Bechtoldc, M. Buschd, B. Szcepaniakd

a Institut für Grenzflächenverfahrenstechnik und Plasmatechnologie, Univ. Stuttgart, D-70569 Stuttgart, Germany b Max-Planck-Institut für Plasmaphysik, EURATOM-IPP, D-85748 Garching, and D-17491 Greifswald, Germany

c NTG Neue Technologie GmbH & Co KG, D-63571 Gelnhausen, Germany

d Galvano-T electroplating-electroforming GmbH, D-51570 Windeck-Rosbach, Germany

At the stellarator Wendelstein7-X (W7-X), an electron cyclotron resonance heating (ECRH) system is

under construction. It consists of 10 gyrotrons operating at 140 GHz, with power up to 1 MW cw each.

The transmission from the tubes to the plasma is performed optically via two multi-beam waveguides [1].

Two switches in this transmission system allow the redirection of the power from two sources to antenna

ports at toroidal positions, where trapped-particle physics and quasi-high-field-side launch can be

investigated [2]. These ports are very narrow and do not allow the installation of front steering launchers,

which are realized for the standard launch of the ten beams from the low-field side. The power will thus

be fed via Remote-steering launchers (RSLs). The remote-steering properties [3] are based on multi-

mode interference in a corrugated square waveguide leading to imaging effects: For a proper

length/width of the waveguide, a microwave beam at the input of the waveguide (with a defined direction

set by a mirror outside of the plasma vacuum) will exit the waveguide near the plasma in the same

direction.

One beam from the multi-mode transmission system is coupled via a steering unit and a diamond vacuum

barrier window into the square corrugated waveguide (length L = 4.6m, width a = 50.2 mm). The

movement of the steering mirror is optimized such that the effective waveguide length is matched to the

angle ϕ of the input and output beams [4]; thus, an extension of the steering range of -15° < ϕ < 15° wrt.

the waveguide axis is anticipated. The corrugation profile of the waveguide was optimized for low loss

and high power capability [5]. The copper waveguide will be manufactured by electro-forming; this

allows a vacuum-tight design with the required precision avoiding deformation by welding, and an easy

integration of the cooling lines. To fit the RSL into the available space at W7-X, a mitre bend is inserted

into the waveguide run; moreover, a gap of 24 mm length is needed to accommodate a vacuum valve.

Both elements require careful optimization to avoid transmission loss and leakage to the valve

mechanism.

The second RSL will be quite similar; however, the steering performance will be further improved by

using an optimized waveguide cross-section [6].

In the paper, design issues and optimization procedures will be presented in detail, calculations on mode

purity as function of the steering angle are shown, and measurements on a mock-up and test samples will

be discussed. The status of fabrication of the RSLs is given. Aspects of this work concerning the

development of reactor-compatible antennas without moveable elements near to the plasma are

addressed.

REFERENCES

[1] V. Erckmann et al, Fusion Science and Technology 52, No2 (2007) 291-312

[2] N. Marushchenko et al., this conference

[3] W. Kasparek et al., “Performance of a remote steering antenna for ECRH/ECCD applications….”. Nucl. Fusion 43 (2003), 505 - 1512.

[4] K. Ohkubo et al., “Extension of steering angle in a square corrugated waveguide antenna”. Fusion Eng. Des. 65 (2003), 657 – 672.

[5] B. Plaum et al., “Numerical calculation of reflection characteristics of grooved surfaces”. J. Infrared Milli. Terahz. Waves 32 (2011), 482–495.

[6] G.G. Denisov et al: “Efficiency enhancement of components based on Talbot effect”. Int. J. Infrared Milli. Waves 28 (2007), 923–935.

This work is supported by the german ministry of education and research (BMBF) by grant 03FUS0017A.

Diamond Window Diagnostics for Nuclear Fusion Applications –

Early Concepts

F. Mazzocchi, G. Aiello, P. Spaeh and T. Scherer

Karlsruhe Institute of Technology, Association KIT-EURATOM, P.O. Box 3640, D-76021 Karlsruhe, Germany

Abstract

The future nuclear fusion power plants will require Electron Cyclotron Heating and Current

Drive (ECH&CD) systems to heat up and stabilize the plasma inside the vacuum vessel. One of

the key components of such systems is the Chemical Vapor Deposition (CVD) Diamond

Window. The purpose of such a device is to act as vacuum and tritium boundary while providing

a nearly perfect microwave transparency. Although perfectly suited for high power microwave

operation, they shall be monitored in order to properly ensure the electron cyclotron heating

system efficiency and safety.

In this paper we show the first assessment study on a set of diagnostics that can be part of a

Diamond Window assembly. The required diagnostics include arc detection, tritium detection,

microwave reflected/transmitted power and disk temperature. The diagnostics devices to be

implemented must have a compact, simple and flexible layout, with a rugged design, in order to

maximize serviceability and durability in the harsh environment they will face. As multiple

options are possible for the various diagnostics systems (e.g. scintillation devices vs solid state

detectors for tritium detection), tradeoffs have been assessed.. In order to accommodate the

additional diagnostics previously mentioned, also a new design for the window housing,

including the copper cuff, was developed.

Next steps foresee the development of an experimental test bench to be inserted in the Launcher

Handling and Test facility (LHT) at KIT, in order to perform validation tests under conditions

similar to the operative ones.

“This work, supported by the European Communities under the contract of Association between

EURATOM and Karlsruhe Institute for Technology, was carried out within the framework of the European

Fusion Development Agreement (GOT4-DIAG WP1). The views and opinions expressed herein do not

necessarily reflect those of the European Commission.”

High Power Experiments of Multi-Frequency Gyrotron

Ryosuke Ikeda, Ken Kajiwara, Yasuhisa Oda, Koji Takahashi and Keishi Sakamoto

Japan Atomic Energy Agency (JAEA), Naka, Ibaraki 311-0193, Japan

An electron cyclotron heating and current drive system in ITER is designed to inject

RF power of 20 MW to control actively plasmas. The RF power is supplied by 24

gyrotrons. The required specifications of a gyrotron are 170 ± 0.3 GHz oscillation

frequency, 1 MW continuous wave operation and 50 % electrical efficiency. High–speed

power modulation of > 1 kHz is required to suppress a neo-classical tearing mode. In

addition, a HE11 mode purity at the MOU exit demands 95 % to achieve low loss

transmission.

In JAEA, the gyrotrons with a triode magnetron injection gun have been developed.

In TE31,8 mode gyrotron, the output power, the electric efficiency and the pulse width

such as 1 MW / 55 % / 800 s and 0.8 MW / 57 % / 3600 s were attained. However, the

marginal power was 1 MW due to increase of heat load in a cavity resonator. Therefore,

In order to reduce the heat load and realize safely more than 1 MW operation, the

gyrotron with high-order mode (TE31,11 mode) is under development. This gyrotron has

advantages for not only heat load in the cavity resonator, but also multi-frequency

oscillation. RF beam of 170 GHz (TE31,11), 137 GHz (TE25,9) and 104 GHz (TE19,7) are

oscillated by varying the magnetic field strength. These beams pass through a diamond

disk window without being reflected and propagate toward the same direction without

moving an final mirror in the gyroton. Measured radiation profiles and the positions at

the front of the window agreed with the calculation results. In short pulse experiments

(< 5s), the output power, the oscillation efficiency and the electric efficiency for 170

GHz, 137 GHz and 104 GHz oscillation are 1.2 MW / 27 % / 43 %, 810 kW / 25 % /

40 % and 720 kW / 25 % / 38 %, respectively. In long pulse experiment of 170 GHz

oscillation, the pulse length reached 100 s with 870 kW output power. The HE11 mode

purity measured at the MOU exit were 95 %, 93 % and 92 % for 170 GHz, 137 GHz

and 104 GHz oscillation. At the exit of 40 m transmission line, the purity of 91 %, 88 %

and 95 %, respectively were obtained. These results and additional progress will be

described.

Research and Development of 2-frequency (110/138 GHz) FADIS for JT-60SA ECH/ECCD Experiments

H. Idei, *S. Moriyama, *T. Kobayashi, *A. Isayama, **M. Sakaguchi, ***W. Kasparek Research Institute for Applied Mechanics, Kyushu Univ., Kasuga, 816-8580, Japan * Japan Atomic Energy Agency, Naka, 311-0193, Japan ** Furukawa C & B Co., Ltd., Yamato, 242-0018, Japan *** Institute of Interfacial Process Engineering and Plasma Technology, Pfaffenwaldring 31, D-70569 Stuttgart, Germany e-mail: idei@kyushu-u.ac.jp FADIS has been developed to switch fast the transmission lines of ECH/ECCD experiments. The local current driven by ECCD has been in operation to suppress NTM mode activities in the tokamak. Synchronous switching for the mode rotation using the FADIS was proposed for more effective suppression. The 2-frequency (110/138 GHz) gyrotron has been developed for a JT-60SA ECH/ECCD system. The FADIS for the 2-frequency gyrotron system is developing under a JAEA collaboration with Kyushu University. First the frequency drift and dip in the gyrotron operation were measured to consider which kind of FADIS is preferred to the JT-60SA ECHCD system. A square corrugated waveguide diplexer system with 2 resonant rings was considered as one of attractive FADIS systems for stable high-power and long-pulse operation in the JT-60SA ECHCD system. Square corrugated waveguide system was proposed for upper port launchers in ITER with remote steering concept. Two kinds of the remote steering launcher were considered. One is symmetric directional launcher whose launching angle was identical for the input, and the other was asymmetric with the opposite directional launching angle for the input, depending on launcher length and side of the square aperture. Symmetric directional launcher with extended launching angle was developed and used for ECH/ECCD experiments on the TRIAM tokamak. The waveguide system has been also considered as a splitter in symmetric and asymmetric directions. The splitter operation is available for the FADIS, but is difficult for the 2-frequancy application in traditional concept. Extended operation region has been surveyed using similar approach for the extended symmetric launcher design. Figure 1 shows Matching Coefficient (MC) being larger than 0.9 between two ideal split beams and the electric field distribution excited at the outlet aperture for the launcher length L and incident angle θ in 2-freqency (110/138 GHz) operation. The side of the aperture was 0.0535 m. High MC operational region was found in normal and 3rd extended branches for 110 and 138 GHz, respectively at θ = 17.5 degree. In the large incident angle of 17.5 degree, the waveguide ohmic loss becomes critical for the high power application. Bell-shaped corrugated structure has been designed with moment-method and FDTD electromagnetic simulators for the large

incident angle. Fabri- cation test has begun to confirm machining accu- racy on the corrugation structure. Fig.1 : Matching Coefficient (MC) for the launcher length L and incident angle θ for 2-freqency (110/138 GHz) operation

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, 22- 25 April 2014, Nara, Japan

Development of a Millimeter-Wave Beam Position and Profile Monitor for Transmission Efficiency Improvement in an ECRH System

T. Shimozuma, S. Kobayashi, S. Ito, Y. Ito, S. Kubo, Y. Yoshimura, M. Nishiura1),

H. Igami, H. Takahashi, Y. Mizuno, K. Okada and T. Mutoh National Institute for Fusion Science, Toki-city, 509-5292, Japan 1)Graduate School of Frontier Science, the University of Tokyo, Chiba-city, 277-8561, Japan e-mail: shimozuma.takashi@LHD.nifs.ac.jp In a high power Electron Cyclotron Resonance Heating (ECRH) system with long-distance transmission lines, the reliable millimeter-wave (mmw) transmission can be much improved by the evacuation, sufficient cooling and precise alignment of the whole transmission system. A real-time beam-position and -profile monitor (BPM) is required to evaluate the position and profile of a high power (Megawatt level) mmw even in the evacuated corrugated waveguide. The idea of the BPM1) is as follows. Two-dimensional array of Peltier devices is installed and aligned on the atmospheric side of a thin miter-bend reflector. A mmw-beam propagating in the corrugated waveguide is reflected on the mirror surface of the miter-bend and partly absorbed in the reflector plate. The generated heat by Ohmic loss diffuses to the outside of the reflector and removed by the Peltier devices. When these devices are connected serially and driven by the constant current control, the voltage change of each device is almost linearly proportional to the temperature change of the cold-side of the device, if the temperature at the hot-side of the Peltier device is kept constant. The information of the two-dimensional temperature profile of the miter-bend reflector can give the real-time information of the position and profile of the mmw beam. If two BPMs are installed apart from about one beat wavelength of the HE11 main mode and the HE21 lowest converted mode in the transmission line, mode contents included could be determined from two beam profile informations2). A higher spatial resolution BPM, which is consisted of 52 Peltier devices with 10 mm square aligned on a circuit board and a water-cooled heat sink, was developed. Initially we tested the BPM using a circular heater with 40mm diameter as a heat source. Algorithm to determine the temperature change from the voltage change of each Peltier device was considered and confirmed experimentally. Additionally, a method of the mode content analysis using the obtained data is considered and proposed. This work is supported by National Institute for Fusion Science under grants ULRR701 and is also supported by a grant for scientific research from the Ministry of Education, Science and Culture of Japan (21560058, 24560066). 1) T. Shimozuma, H. Takahashi, S. Ito, et al., Proceedings on 36th International

Conference on Infrared, Millimeter, and Terahertz Waves, 2011, Houston, TX, USA W2A-1.

2) T. Shimozuma, S. Kobayashi, S. Ito, et al., Plasma Conference 2011, Nov. 22-25, Kanazawa, Japan, 22P146-P.

Development of power/polarization monitor

in the ECRH transmission line

R. Makinoa, S. Kuboa,b, K. Kobayashia, S. Kobayashib, T. Shimozumab, Y. Yoshimurab,

M. Nishiurab, H. Igamib, H. Takahashib, S. Ogasawaraa, T. Idoa,b, T. Mutohb

a Department of Energy Engineering and Science, Nagoya Univ., Nagoya 404-8603

b National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292

makino.ryohhei@ms.nifs.ac.jp

For optimization of electron cyclotron resonance heating (ECRH), it is important to measure

power and polarization states of injected millimeter-waves. Arbitrary polarization states of

millimeter-waves can be realized by two grating (λ/4 and λ/8) mirror polarizers set at

miter-bends in the corrugated waveguide transmission system on the Large Helical Device

(LHD). The polarization state of an injected millimeter-wave determines the mode excitation

purity, and therefore the power absorption efficiency in plasmas. The development of

real-time monitor that enables simultaneous detection of the injected power and polarization

state is required.

The power/polarization is measured by the monitor composed of bi-linear polarization

directional coupler and heterodyne interferometer that enables to detect the power and the

phase difference of two orthogonal polarizations. IF signals are measured by fast ADC which

has sampling rate of 800 MHz with FPGA (Field Programmable Gate Array). The power and

polarization state of EC waves can be estimated by measuring phases and amplitudes of both

polarizations simultaneously. The power/polarization monitor was set on a miter-bend on a

transmission line of millimeter-waves. 77 GHz EC waves were injected to the transmission

line. Grating polarizers were rotated to change polarization states of the millimeter-waves.

The phases and amplitudes of E-polarization and H-polarization were detected. The results

agreed with the theoretical values qualitatively.

Recent results of gyrotron operation in NIFS

S. Ito, T. Shimozuma, Y. Yoshimura, H. Igami, H. Takahashi, M. Nishiura1), S. Kobayashi, Y. Mizuno, K. Okada, S. Kubo

National Institute for Fusion Science, Toki, Japan

1)Graduate school of the University of Tokyo, Japan

e-mail:ito.satoshi@nifs.ac.jp

In the last LHD experimental campaign, a 154GHz gyrotron which had been conditioned to generate 1 MW/2 s, 0.5 MW/CW was installed for LHD experiments. Four high power gyrotrons (three-77 GHz/1~1.5 MW and one-154 GHz/1 MW) and a CW gyrotron (84 GHz/0.2 MW) are ready. Our experimental group tends to need high energy and various injection patterns to ECRH recently. Higher individual injection power and various injection patterns, we developed power enhancement method by stepped anode acceleration voltage control and tried to operate the gyrotron in hard excitation region. These operations are realized by remote controlled waveform generator. But oscillation map of high power or long pulse operation in the hard excitation region were limited, because in order to achieve hard excitation region by anode voltage control, one need to pass through high anode current phase within a time short enough that anode or anode power supply do not be overloaded. This limitation becomes more critical when the gyrotron beam current is increased to increase the output power. In the long pulse operation, it was impossible to reach the hard excitation region in low beam current (<10A). We will present recent condition of these gyrotrons in high power/short pulse and low power/CW regimes. Some other development of ECRH related components like 1MW/CW dummy load are also discussed.

Digital Correlation ECE Measurement Technique with a Giga Hertz Sampling Digitizer

H. Tsuchiya, S. Inagakia, T. Tokuzawa, N. Tamura, Y. Nagayama

National Institute for Fusion Science,322-6 Oroshi-cho, Toki 509-5292, Japan aResearch Institute for Applied Mechanics, Kyushu University, Kasuga 816-8580, Japan

tsuchiya.hayato@lhd.nifs.ac.jp

In magnetized plasmas, it is widely recognized that turbulence phenomena as well as MHD instability are considered to have an impact on the confinement characteristic, and the understanding of turbulence structure is a quite very important issue to establish a discharge scenario with better confinement. For gaining a better understanding of the physical mechanism of the relationship between the turbulence and the transport, it is essential to observe spatiotemporal structures of micro- to macro-scale turbulence at the same time. In this contribution, we propose a Digital Correlation Electron Cyclotron Emission (DCECE) measurement technique for observing the micro-scale electron temperature fluctuation in LHD. In the conventional ECE measurement system, including a typical correlation ECE measurement system, band pass filters are usually used in an intermediate frequency (IF) band that is down-converted from a radio frequency band. The frequency band of the filters determines the spatial resolution, and thus it is unmodifiable after the data has been obtained. A technique of digitizing IF band (the frequency range of that is several Giga Hertz) is applied to the correlation ECE, with a Giga Hertz Sampling Digitizer. Because the IF spectra is obtained by FFT analysis of digitized IF wave form data, one can choose the temporal and spatial resolution even after the date has been acquired. This is the greatest merit of this method. In order to improve the sensitivity limit, it should be necessary that the Te fluctuation is analyzed by the same process of the conventional correlation ECE. The result of the analysis of test wave and the initial analysis of the data obtained in LHD will be discussed at the meeting.

Overview and prospective of the ECE measurements on EAST

Y. Liu1*, X. Han1, X. Liu1, A. Ti 1, E. Z. Li1, C. W. Domier2, N.C. Luhmann, Jr.2, J. Fessey3, P. Trimble3, S. Schmuck3, M. Xu1, B.L. Ling1, L. Q. Hu1, B. N. Wan1, and J. G. Li1

1 Institute of plasma physics, Chinese Academy of Sciences, Hefei 230031, China 2 Department of Applied Science, University of California at Davis, Davis, California 95616 3 Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon. OX14 3DB, United Kingdom * liuyong@ipp.ac.cn

ECE measurements on EAST have experienced massive changes in 2013. In

contrast to the previous situation where individual transmission line was used for

each system, new transmission lines for horizontal/vertical line of sight were

constructed and shared with different systems for the same objective. The horizontal

transmission line is composed of quasi-optics, ~45 meters corrugated waveguides, 8

miter bends, and a power divider. The quasi-optics form a beam pattern of roughly 5

cm FWHM. The transmission loss is estimated to be below 3 dB.

In total there will be four systems operating in 2014 experimental campaign. Two

systems including a 32-channel heterodyne radiometer and a Michelson

interferometer are for electron temperature profile measurement along a horizontal

chord. The other two systems are a 16-channel heterodyne radiometer and a

20-channel grating polychromator, and they measure the ECE along a vertical chord

for monitoring the generation of suprathermal electrons qualitatively. These systems

provide a comprehensive ECE measurement on EAST with fairly good temporal

and spatial resolution for a wide toroidal magnetic field range. In order to provide

independent absolute electron temperature information, the systems are calibrated

by using the hot/cold load method.

In addition to the four existing systems, a correlation ECE will be designed in 2014

for the measurement of electron temperature fluctuations in plasma core region.

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating

Multiple flux tubes driven by ECH on or inside q=1 surface of

sawtoothing KSTAR plasmas†

G. S. Yun1*

, A. Bierwage2, G. H. Choe

1, Y. B. Nam

1, M. J. Choi

1, W. Lee

1,

H. K. Park3, Y. S. Bae

4, N.C. Luhmann Jr.

5

1Pohang University of Science and Technology, Pohang, Gyungbuk 790-784, Korea

2Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212, Japan

3Ulsan National Institute of Science and Technology, Ulsan 689-798, Korea

4National Fusion Research Institute, Daejeon 305-333, Korea

5University of California, Davis, CA 95616, U.S.A

*E-mail: gunsu@postech.ac.kr

In experiments on the KSTAR tokamak to control the internal sawtooth instability by electron

cyclotron resonance heating (ECH), multiple flux tubes of m/n=1 helicity have been routinely

observed on or inside the q=1 surface during the inter-crash period by a 2-D electron cyclotron

emission imaging (ECEI) diagnostic system [1]. The initial location of the flux tubes coincides

with the ECH position and the number of flux tubes increases as the ECH position approaches

the q=1 surface [2]. The flux tubes later merge into a single m=1 flux tube, which eventually

crashes like the usual sawtooth crash, i.e., crash of an internal kink mode. The topology and

merging of the flux tubes suggest that the flux tubes are initially generated by ECH, presumably

in a flattened q profile by the preceding sawtooth crash, and grow with a nonlinear interaction

with ECH, carrying a large fraction of the current driven by ECH. A reduced MHD simulation

[3] including a source term modeling the time-varying interaction between the ECH beam and

flux tubes will be described and the initial results will be compared with the observations. †Work

supported by NRF of Korea and the U.S. DoE. [1] G. S. Yun et al., Phys. Rev. Lett. 109, 145003

(2012). [2] G. Choe et al., 9th

Asia-Pacific Fusion Association Conference (2013). [3] A.

Bierwage, S. Hamaguchi, M. Wakatani, S. Benkadda, and X. Leoncini, Phys. Rev. Lett. 94,

065001 (2005).

Figure

(Left) Model for helical flux

tubes formed by ECH.

(Right) Observed dual flux

tubes by ECEI. The dotted

circle is the inversion radius

estimated from the crash.

Detection of MHD instabilities with ECE

Hugo van den Brand, M.R. de Baar, M. van Berkel, W.A. Bongers, N. Doelman, L. Giannone,

W. Kasparek, J.K. Stober, D. Wagner, E. Westerhof and the ASDEX Upgrade team

January 31, 2014

The viable operation of a fusion reactor requires suppression of βN -limitingMagnetohydrodynamic (MHD) instabilities. Neoclassical Tearing Modes (NTMs)are limiting in particular causing a reduced risk of disruption and at least a dropof core temperature. Direct detection of the position of this instability is in-vestigated with inline Electron Cyclotron Emission (ECE), in which ECE isdetected via the transmission line of the EC heating system. The progress ofNTM detection using inline ECE on ASDEX Upgrade will be discussed.

Alternatively, NTMs can be prevented by reducing the sawtooth amplitude.This can be achieved by maintaining a small sawtooth period. Methods for thedetection of the sawtooth period will be discussed.

Session: ECE

Preference: Oral

1

Absolute intensity calibration of ECE measurements for electron temperature

measurement on EAST

X. Liu1, Y. Liu1*, E. Z. Li1, X. Han1, J. Fessey2, P. Trimble2, S. Schmuck2, A. Ti1, M. Xu1, and L. Q. Hu1

1 Institute of plasma physics, Chinese Academy of Sciences, Hefei 230031, China 2 Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon. OX14 3DB, United Kingdom

* liuyong@ipp.ac.cn

Two ECE measurement systems have been commissioned on EAST for electron

temperature profile measurement, and they are a 32-channel heterodyne radiometer

and a Michelson interferometer. An in-situ calibration system was incorporated into

the transmission line by using a rotatable ellipsoidal mirror. The hot/cold load

method is utilized to calibrate the systems. The hot source borrowed from JET can

be heated up to 600 Celsius degree, and the cold load is either the ambiance or the

absorber immersed in LN2. To detector the ultra-weak signal from the calibration

source, a lock-in amplifier and a chopper are employed for the calibration of the

radiometer. On the other hand, the outputs are acquired directly for long time to do

coherent average and this improves the signal to noise ratio. These two methods

cross-checked the reliability of the absolute calibration. Regarding the calibration of

the Michelson system, it is relatively easier benefiting from the high light

throughput of the system. Thousands of interferograms are acquired and averaged

coherently.

A study of mode purity improvement in the ITER relevant transmission line

Y. Oda, R. Ikeda, K. Kajiwara, K. Takahashi, K. Sakamoto

Japan Atomic Energy Agency

In JAEA, ITER relevant transmission line (TL) was examined using a 170

GHz high power gyrotron. In this paper, the activity to improve of mode

purity in TL is reported.

The TL test stand in JAEA is composed of 63.5 mm diameter corrugated

waveguide(WG) system. The TL test stand delivers 170 GHz high power RF

from the gyrotron to dummy loads or the ITER equatorial port launcher (EL)

mockup. The length of long distance TL which includes 4 miter bends, a

couple of polarizer, and two WG switch was 40 m. The RF power from

gyrotron was coupled into TL inlet at a matching optical unit (MOU) with

95% of LP01 mode purity.

While assembly of the TL test stand, the mode content of input and end of

the straight section was measured. Then 5% of mode purity degradation was

found in the 12 m length line though it was just assembly of 2m straight

WGs. Indeed, the mode conversion calculation showed that 5% of mode

conversion may occur when the straight section causes 0.5 mm of periodic

deflection. To minimize the deflection, the alignment method with

adjustment of the position and angle of the WG piece utilizing the laser beam

reflection was applied for the TL re-assembling. The mode purity in the same

section was improved and mode purity degradation was less than 1%. Finally

91% of LP01 mode purity was achieved at the end of 40 m length TL.

The mode conversion caused in during the long pulse operation of TL test

stand was measured. While TL delivers high power RF from the gyrotron

during long pulse operation and TL components are heated. The thermal

expansion of the long section deforms neighboring sections and this increase

the mode conversion in TL. The effect of long pulse operation on mode purity

will be discussed.

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, 22-

25 April 2014, Nara, Japan

Status of Europe’s contribution to the ITER EC

system

F. Albajar1, T. Bonicelli1, G. Carannante1, M. Cavinato1, F. Cismondi1, C. Darbos,

P. Dharmesh2, M. Gagliardi1, F. Gandini

2, T. Gassmann

2, M. Henderson

2,

R. Nousiainen1, G. Saibene1, F. Sartori1, K. Takahashi2, and the EU Team

1Fusion for Energy, Josep Pla 2, Barcelona, 08019, Spain

2ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance,

France

The EC system of ITER for the initial configuration will provide 20MW of RF power

into the plasma during 3600s and a duty cycle of up to 25% for heating and (co and

counter) current drive generation, also used to control the MHD plasma instabilities.

The system is comprised of 12 high voltage power supplies (HVPS), 24 microwave

sources, 24 transmission lines (TL), 1 equatorial launcher (EL), 4 upper launchers

(UL) and an integrated main control system (MS). This system is being procured by 5

domestic agencies, which includes Europe (F4E), plus the ITER Organization (IO).

F4E has the largest fraction of the EC procurements, which includes 8 HVPSs, 6

microwave sources, the ex-vessel waveguides (includes isolation valves and diamond

windows) for all launchers, 4 upper launchers and the Main Control system. The large

contribution covers nearly the entire spectrum of the EC system from plug to plasma.

F4E is working with IO to improve the overall design of the EC system, which

includes integration of technological advances, simplification of interfaces, global

engineering analysis and assessment of EC H&CD physics and technology

capabilities. Examples are the optimization of the HVPS and gyrotron requirements

and performance relative to power modulation for MHD control, common

qualification programs for diamond window procurements, and assessment of the EC

grounding system. This paper aims at providing a status of Europe’s contribution to

the ITER EC system, and a summary of the global activities underway by F4E in

collaboration with IO for the optimization of the subsystems.

Acknowledgments. This work is carried out in collaboration with the IO Organization. The views

expressed in this publication are the sole responsibility of the author and do not necessarily reflect the

views of the ITER Organization and Fusion for Energy. Neither Fusion for Energy nor any person

acting on behalf of Fusion for Energy is responsible for the use which might be made of the

information in this publication.

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, 22-25 April 2014, Nara, Japan

Status and progress of the European ITER gyrotron development

G. Gantenbein1, F. Albajar2, K. Avramidis1, A. Bruschi3, F. Cismondi2, J.-P. Hogge4, S. Illy1, J. Jelonnek1, J. Jin1, I. Gr. Pagonakis1, Y. Rozier5, T. Rzesnicki1, A. Schlaich1, M. Thumm1, I. G. Tigelis6

1Karlsruhe Institute of Technology (KIT ), Institute for Pulsed Power and Microwave Technology (IHM), Association EURATOM-KIT, Kaiserstr. 12, 76131 Karlsruhe, Germany 2European Joint Undertaking for ITER and the Development of Fusion Energy (F4E), Barcelona, E-08019, Spain. 3 Istituto di Fisica del Plasma, Consiglio Nazionale delle Ricerche, EURATOM-ENEA-CNR Association, 20125 Milano, Italy 4 Centre de Recherches en Physiques des Plasmas, Association EURATOM-Confédération Suisse, EPFL, Ecublens, Lausanne, CH-1015, Switzerland. 5 Thales Electron Devices, 2 rue Marcel Dassault, Vélizy-Villacoublay, F-78141, France. 6 Faculty of Physics, National and Kapodistrian University of Athens, Zografou, GR-157 84, Athens, Greece. e-mail: gerd.gantenbein@kit.edu

Abstract

KIT is strongly involved in the EU development of 170 GHz high power gyrotrons for the ECH&CD system of the ITER tokamak. The focus in this project is on a day-one conventional 1 MW continuous wave (CW) gyrotron which is performed within the European EGYC Consortium and an advanced coaxial 2 MW gyrotron which is planned for later upgrade and power enhancement of the ITER system. The scientific design of major components of the 1 MW tube, e.g. magnetron injection gun, beam tunnel, cavity, quasioptical output coupler and single-stage depressed collector, is based on the experience and improvements achieved with the series production of the 140 GHz, 1 MW, CW gyrotrons for the stellarator W7-X. In collaboration with the EU industrial partner Thales Electron Devices, the industrial design of the technological parts of the gyrotron has been completed. In order to validate the physics design of the 170 GHz, 1 MW CW gyrotron a short pulse (SP) tube with a minimum of cooling technology allowing a pulse length of approximately up to 10 ms is under manufacturing. This tube will allow a flexible and fast replacement and examination of components during the development phase. The design of the tube, the status of procurement and assembling of the SP gyrotron and the foreseen test procedure at the KIT test facility will be presented. Since coaxial-cavity technology has the potential of considerably higher output power, KIT is persuing this development, as a very attractive solution for the ECH&CD systems for future fusion reactors. Due to the very high output power and high electron beam current density one of the possible issues of the coaxial design is the excitation of parasitic oscillations which reduce the efficiency of the system and may result in internal power deposition at unfavorable locations which may suffer from thermal overload. At KIT low frequency parasitic oscillations in the MHz range and electron beam instabilities have been studied, both experimentally and theoretically. Several improvements have been included in the experimental set-up of the short pulse gyrotron and stable operation with an RF output power at 2.3 MW has been achieved.

Acknowledgment This work, supported by the European Communities under the contract of association between EURATOM

and KIT, was carried out within the framework of the European Fusion Development Agreement. Part of this work is supported by Fusion for Energy under Grants F4E-GRT-432 and OPE-458 and within the European GYrotron Consortium (EGYC). EGYC is a collaboration among CRPP, Switzerland; KIT, Germany; HELLAS, Greece; IFP-CNR, Italy. The views expressed in this publication do not necessarily reflect the views of the European Commission.

Recent Tests on 117 GHz and 170 GHz Gyrotrons

K. Felch, M. Blank, P. Borchard, S. Cauffman Communications and Power Industries, 607 Hansen Way, Palo Alto, CA 94304, USA

Two megawatt-class gyrotrons at frequencies of 117 GHz and 170 GHz have recently been

fabricated and tested at CPI. The 117 GHz gyrotron was designed to produce up to 1.8 MW at 117.5 GHz, for 10-second pulses, and will be used for electron cyclotron heating and current drive on the DIII-D tokamak at General Atomics. The 170 GHz gyrotron is specified as a 500 kW CW system, but has been designed with the goal of generating up to 1 MW CW. Oak Ridge National Laboratory will use the gyrotron in ITER ECH transmission line testing.

The 117 GHz gyrotron employs a depressed collector, a diode magnetron injection gun, a cylindrical interaction cavity in which the TE20,9,1 mode is excited, and an internal converter to transform the excited mode into a high-quality fundamental Gaussian output beam, which exits the gyrotron horizontally through a CVD diamond output window. The collector is constructed from a strengthened copper alloy to mitigate the effects of cyclic fatigue.

Initial testing of the gyrotron achieved all performance goals except for the ability to operate at extended pulse lengths. The short-pulse (~5 ms) tests demonstrated output power levels up to 1.8 MW (for a beam current of 60 A, an accelerating voltage of 98 kV, and a collector depression voltage of 25 kV). Output power levels up to 1.5 MW were obtained with a beam current of 50 A. Operation in the desired mode was observed over a broad range of operating parameters with little evidence of mode competition. Internal diffraction losses were measured calorimetrically to be about 3.5% of the total output power. Cavity power losses were as predicted, and no excess power was absorbed in the beam tunnel. Collector power distribution measurements using external temperature sensors confirmed that the spent electron beam power was spread over a broad range of the collector surface, and that the peak time-averaged power densities were within acceptable limits. Thermal imaging of the output beam confirmed the proper operation of the internal converter, consistent with cold-test measurements that were performed prior to completion of gyrotron assembly. Work is currently underway to improve the long-pulse performance of the gyrotron.

The 170 GHz gyrotron employs an electron gun with a modulating anode, to provide separate control of the beam radius, pitch factor, and voltage. The nominal design point for generating 1 MW of output power is an accelerating voltage of 75 kV (80 kV max) and a beam current of 45 A (50 A max). A depressed collector, with a nominal depression voltage of 27 kV (30 kV max) is employed. An edge-cooled CVD diamond output window is used to transmit the gyrotron’s Gaussian output beam with minimal loss. The electron beam produced by the gun interacts with the TE31,8,1 mode of the interaction cavity, and this mode is converted to a Gaussian output beam using an internal converter consisting of a dimpled-wall launcher and three phase-correcting mirrors. Like the 117 GHz gyrotron, the 170 GHz gyrotron collector employs a strengthened copper alloy, as well as iron beam shaping and active magnetic sweeping to reduce both instantaneous and time-averaged power densities to levels compatible with long life.

Factory testing of the prototype 170 GHz gyrotron was recently completed. Several factors hampered the ability to achieve operation at the intended high-efficiency region of parameter space. Access to the highest-efficiency operating regime was limited by mode competition, which was determined to be a result of the mod-anode gun startup scenario. In addition, operation at the lower magnetic fields expected to yield the highest efficiency also resulted in sporadic mod-anode current, and the presence of even very low levels of mod-anode current prevented long-pulse operation from being reliably achievable. Finally, absorption of power in the beam tunnel was found to increase with operating voltage, indicating that the absorbing material in the beam tunnel was not sufficient to prevent the excitation of unwanted beam tunnel oscillations. As a result of these various factors, the maximum output power achieved during initial tests was limited to 600 kW for short pulses and 300 kW for long (15 second) pulses. Nevertheless, these tests did allow for confirmation of the optimal choice of electron beam radius for excitation of the desired mode. The gyrotron has been rebuilt with a diode electron gun and a new beam tunnel design. Test results on the rebuilt 170 GHz gyrotron will be presented if they are available.

18th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, 22-25 April 2014, Nara, Japan

From W7-X Towards a Gyrotron Design for DEMO: Ongoing Research and Development at KIT

J. Jelonnek, K. A. Avramidis, J. Franck, G. Gantenbein, K. Hesch, S. Illy, J. Jin, A. Malygin, I. Gr. Pagonakis, T. Rzesnicki, A. Samartsev, A. Schlaich, T. Scherer, D. Strauss, M. Thumm, J. Zhang

Karlsruhe Institute of Technology (KIT), Association EURATOM-KIT, Kaiserstr. 12, D-76131 Karlsruhe, Germany e-mail: john.jelonnek@kit.edu

Abstract

Today’s requirements for W7-X series gyrotrons are 1 MW RF output power at an operating frequency of 140 GHz. The required efficiency at CW operation is 45 %. For its initial installation phase, ITER is requesting 170 GHz, 1 MW, CW gyrotrons with an efficiency of about 50 %. If comparing this step of development from W7-X to ITER with the requirement for a future demonstration power plant (DEMO), it is relatively moderate. Current designs for DEMO demand frequencies above 230 GHz for efficient ECCD, and, a gyrotron efficiency of better than 60 % to achieve a proper fusion gain factor. Considering the total required EC heating power for DEMO, multi-MW gyrotrons with a unit power clearly above 1 MW will be required, additionally. Fast frequency tuning in steps of around 2-3 GHz is requested. At last, operation in leaps of about 30 – 40 GHz is considered advantageous for plasma heating and control. The combination of these requirements for DEMO clearly challenges present day technological limits. The finally achievable output power at the required operating frequency above 230 GHz will be determined by the continuously evolving boundaries of the core technologies, which are the maximum thermionic emission for long emitter lifetimes, the maximum thermal wall loading, the achievable collector efficiency and, finally, the required tolerances in manufacturing.

Design studies towards DEMO-compatible gyrotrons and related components technologies have been started. They range from theoretical studies on the proper mode selection, studies for cavity and magnetron injection gun to experimental studies on multi-frequency operation of gyrotrons and emitter technologies. In this frame, a novel mode-selection strategy for coaxial gyrotrons operating with high-order modes (eigenvalue χ~150) has been developed. The new strategy focuses on mode restrictions imposed by the assumed quasi-optical launcher and single-disk window of the gyrotron rather than on spectral considerations, since the mode spectrum in such gyrotron cavities is nearly independent from the actual main operating mode.

The presentation shall provide a comprehensive overview on the research which shall lead to DEMO-compatible gyrotrons at KIT.

Acknowledgment

This work, supported by the European Communities under the contract of association between EURATOM and KIT, was carried out within the framework of the European Fusion Development Agreement. The views expressed in this publication do not necessarily reflect the views of the European Commission.

Technology

Development in Russia of Gyrotrons for Fusion. Status and New Trends.

G.G.Denisov and IAP/GYCOM team

Institute of Applied Physics Russian Academy of Sciences GYCOM Ltd

During last years several new gyrotrons were developed at IAP/GYCOM. Main efforts were spent for development 170GHz/1MW/50%/CW gyrotron for ITER and multi-frequency gyrotrons. The main ITER requirements to a gyrotron have been demonstrated: 1MW power, 1000 seconds pulse duration, 53% efficiency. The gyrotron operation regime of 1.2 MW was found for 100 second pulses. Now the main activity for ITER gyrotron is enhancement of their reliability. For a multi-frequency gyrotron a novel scheme for a tuneable window is studied. There are also two new directions in the gyrotron development: elaboration of gyrotrons at higher frequencies (250-300GHz) and investigation of the possibility to realize frequency locking, phase stabilization, efficiency increase by means of an external signal. In fact these two problems can be linked. Megawatt-power gyrotrons with high transverse indices of an operating mode have very dense spectrum of Eigenmodes. Single-mode generation with high efficiency and stable fre-quency is hard to achieve due to some reasons. Firstly, the optimal operating point is in the hard-self excitation region. Secondly, gyrotron startup scenario begins with noise-level ampli-tudes of all modes, so there is a certain possibility to get a wrong generation or multimode generation by the end of turn-on process. Also the power supply voltage instability results in the frequency instability and the latter blocks some gyrotron applications. One of the ways to overcome these problems is synchronizing the gyrotron by an external signal of constant fre-quency. If this signal has enough power with respect to the power of single-mode generation at optimal parameters without the signal, there is a definite frequency band where frequency and phase of the work mode can be captured by the signal. As a result, work mode stabilizes and suppresses parasitic modes and can be driven to the maximal efficiency point through the startup scenario. The proof-of-principle experiment on the phase locking is in progress.

Preliminary design of powerful gyrotrons for IGNITOR and DEMO V.E.Zapevalov, A.V.Chirkov, G.G.Denisov, A.N.Kuftin, A.G.Litvak, M.A.Moiseev, N.A.Zavolsky

Institute of Applied Physics, Russian Academy of Sciences, Nizhny Novgorod, Russia

zapev@appl.sci-nnov.ru,

For some advanced fusion experiments at the next plasmas like IGNITOR and DEMO and promising techno-

logical applications RF sources with CW power near 200-1000 kW and frequency 240-300 GHz are requested

[1]. In the framework of this project, the program for design of a 300 GHz gyrotron with high output power (up

to 1000 kW) for operation in the continuous regime was implemented at IAP RAS.

Gyrotron tube version with 200 kW output power could to be used together with 12 T liquid Helium free cry-

omagnet of SHI or JASTEC Company (Japan) [2] or its analogue with diameter of a “warm” aperture 100 mm.

On the basis of the analysis of scientific and technical information the general design concept of the gyrotron has

been developed and possible operating modes and the value of accelerating voltage have been chosen. Several

possible operating modes were considered. Finally on the basis of the analysis of technical limitations the TE22,8

operating mode have been chosen for 200 kW gyrotron tube version which will give us the possibility to check

different scientific and technical solution. Accelerating voltages for this tube is 60 kV and beam current 20 A.

The gyrotron tubes have the same principal design concept as the earlier developed 110-170 GHz gyrotrons [1]

and in part the 300 GHz/4kW/CW gyrotron [2]. This includes the use of a quasi-diode gun with an impregnated

or LaB6 cathode, a build-in mode converter to transform cavity mode to wave beam with radial directed output,

and an electron beam collector, all designed for CW operation.

Using potential depressed collector (CPD) give us the possibility really improve gyrotron efficiency. Effi-

ciency without CPD achieves 23.05 % without after-cavity interaction (ACI) [3], and 22.12 % with ACI. Effi-

ciency with CPD and 0 % electron reflection is 39.6%, and with CPD and 1 % reflections 42.8 %. The mode

converter includes a specially shaped waveguide end and three profiled mirrors to provide: low diffraction losses

inside the tube, optimal RF power distribution over an output window; matching of the output wave beam to a

transmission line. The TEM00 mode content at the tuber output is 99.54 %. For the output window of

300 GHz/200 kW CW gyrotron will be used the CVD-diamond with near 80 mm diameter disks with small RF

loses.

More powerful 300 GHz tube version needs 13 T cryomagnet with increased diameter of a “warm” aperture

accelerating voltages near 80 kV and beam current 40-50 A. Particular properties of the 240 GHz gyrotron for

DEMO were considered also. Results of this activity are summarized in this report.

References

1. A.G.Litvak, G.G.Denisov, M.V.Agapova, et al. Recent Results of Development in Russia of 170 GHz Gyro-

tron for ITER // The 35-th Int. Conference on Infrared, Millimeter and Terahertz Waves, 2010, Sept.5- Sept.10

Roma, Italy, Conference Digest, p.Tu.-E1.1

2. V.Bratman, M.Glyavin, T.Idehara, et al. Review of Sub-Terahertz and Terahertz Gyrodevices at IAP RAS

and FIR FU // IEEE Transactions on Plasma Science, 37, 1, P. 36-43, (2009)

3. V.E.Zapevalov. Increasing Power and Efficiency of gyrotrons. Fusion Science and Technology, August

2007 Vol.52, No2, P. 340-344

Development of a dual frequency (110/138 GHz) gyrotron for JT-60SA

and its extension to an oscillation at 82 GHz

T. Kobayashi, S. Moriyama, A. Isayama, M. Sawahata, M. Terakado, S. Hiranai, K. Wada,

Y. Sato, J. Hinata, K. Yokokura, K. Hoshino, and K. Sakamoto

Japan Atomic Energy Agency, Naka, 311-0193, Japan

e-mail: kobayashi.takayuki@jaea.go.jp

In JT-60SA, an electron cyclotron heating and current drive (ECH/CD) system will have

nine gyrotrons with the total injection power of 7 MW and the maximum pulse duration of

100 s [1]. The main role of the ECH/CD system is localized ECH and ECCD for sustainment

of high-beta plasma.

A dual-frequency gyrotron, which can generate 110 GHz and 138 GHz waves

independently, is being developed in JAEA [2] to enable ECH/CD in a wider range of plasma

discharge conditions. The target output power and pulse length are 1 MW and 100 s,

respectively, at both frequencies and oscillations of 1 MW for 10 s at both frequencies were

obtained, so far. The 110 GHz wave will be useful for ECH/CD at the toroidal magnetic field

Bt of ~ 1.7 T, while the 138 GHz wave will be useful at Bt ~ 2.3 T [3]. Both of them are

injected as second harmonic resonance waves in JT-60SA. In gyrotron design, it is essential to

choose an operating TEmn mode desired for the target frequency. Key parameters to obtain

high-power, long-pulse and stable oscillations without mode competition are the cut off

frequency, fc = 'mnc/2a, the thickness of the output window, dwindow = N×w/2, the caustic

angle in the quasi-optical mode convertor, = 2 cos('mn/m), and the beam radius in the cavity

rb = 'm-1,1 /2. Here, 'mn, c, a, N, and w are the n-th root of the derivative of m-th order

Bessel function, the speed of light, the cavity radius, integer, the wavelength in vacuum, and

the wavelength in the window material (CVD diamond), respectively. In the case of

dual/multi frequency gyrotrons, selection of operating modes is one of the most important

issues, in which all of the above parameters should be satisfied simultaneously for dual/multi

frequency. In the design of the dual frequency gyrotron for JT-60SA [3], we found that

operating modes of TE22,8 for 110 GHz and TE27,10 for 138 GHz are the best combination to

achieve high-power (> 1 MW) and long pulse of (> 100 s).

In addition to the above two frequencies, we considered a possibility of an oscillation at

82 GHz as an additional frequency of the dual frequency gyrotron in design. We found that

the oscillating mode of TE17,6 at 82 GHz satisfies the above conditions except slightly large

difference in the caustic radius. In calculation, an oscillation power of 1 MW is obtained at

the beam current of > 50 A with the acceleration voltage of 80 kV and the electron pitch

factor of 1.1 at 82 GHz. The calculated diffraction loss in the gyrotron is around 6% and it is

sufficiently low for short pulse oscillation of < 1 s at < 1 MW. In 2013, we operated the

gyrotron at the cavity field of ~3.3 T, and an oscillation at 82 GHz was demonstrated with the

output power of 300 kW and the pulse length of 20 ms. The pulse length was limited by the

capability of a small dummy load used in this experiment. The burn pattern at the output

window and the measured frequency were agreed with the calculation. Higher power and

longer pulse length operation will be carried out in the future. Since the 82 GHz wave can be

injected as a fundamental resonance harmonic wave at Bt ~ 2.3 T, there is a possibility of the

use of this frequency for effective start-up assist and ECH wall cleaning.

References

[1] Y. Kamada et al., Nucl. Fusion 53 (2013) 104010.

[2] T. Kobayashi et al., Trans. Fusion Sci. Technol. 63, 1T(2013) 160.

[3] A. Isayama et al., Plasma Fusion Res. 7 (2012) 2405029.

Development of Resonant Diplexers for high-power ECRH -

Status, Applications, Plans

W. Kasparek

1, B. Plaum

1, C. Lechte

1, Z. Wu

1, H. Wang

1, M. Maraschek

2, J. Stober

2, D. Wagner

2, M.

Schubert2, G. Grünwald

2, F. Monaco

2, S. Müller

2, H. Schütz

2, V. Erckmann

2, N. Doelman

3, R. van den

Braber3, W. Klop

3, H. van den Brand

4, W. Bongers

4, B. Krijger

4, M. Petelin

5, L. Koposova

5, L.

Lubyako5, A. Bruschi

6, K. Sakamoto

7, teams at the contributing institutes, and ASDEX Upgrade Team.

1 Institut f. Grenzflächenverfahrenstechnik und Plasmatechnologie, D-70569 Stuttgart, Germany

2 Max-Planck-Institut für Plasmaphysik, EURATOM Ass., 85748 Garching and 17491 Greifswald, Germany

3 Department of OptoMechatronics, TNO Tech. Sciences, NL-2600 AD, Delft, The Netherlands

4 Dutch Institute For Fundamental Energy Research, NL-3439 Nieuwegein, The Netherlands

5 Inst. of Applied Physics, Russian Academy of Science, 603950 Nizhny Novgorod, Russia

6 Istituto di Fisica del Plasma, EURATOM-ENEA-CNR Ass., I-20125 Milano, Italy

7 Japan Atomic Energy Agency (JAEA), 801-1, Mukoyama, Naka, Ibaraki 311-0193, Japan

The development of diplexers for ECRH has been pursued at a number of institutes because

of their attractive variety of applications: Power combination, non-mechanical, electrically

controlled switching (of combined beams) between launchers with tens of kHz, and

discrimination of low-power ECE signals from high-power ECRH is feasible. Especially ring

resonators have demonstrated high performance; therefore, several devices of this type

operating at different frequencies and waveguide standards have been built and investigated.

The status of development, recent results from experiments at ASDEX Upgrade and near-term

plans are given in this paper.

The resonant diplexer Mk IIa, which employs a resonator for Gaussian beams and mode

converting horns to match the connecting HE11 waveguides, was tested extensively at the 140

GHz, W7-X ECRH system [1]. The device was equipped with a frequency-tracking system to

compensate drifts of the gyrotron frequency, and is now installed in the ASDEX Upgrade

ECRH system [2]. Commissioning experiments on fast switching between two launchers for

synchronous stabilization of neoclassical tearing modes, as well as in-line ECE measurements

have been performed successfully. Experimental issues and first results are discussed. A clear

influence of the switching phase on the amplitude of the 3/2 NTM mode was measured,

complete stabilization could, however, not be demonstrated yet due to lack of power and

imperfect resonator control. Results from in-line ECE experiments from ASDEX-Upgrade are

presented in Ref. [3].

Present diplexer versions feature direct corrugated waveguide inputs, owing to the use of ring

resonators with phase-reversing mirrors matched to HE11 fields. Low-power measurements

of such a diplexer (MC IIIb) show high performance, close to theory. Because of the compact

design, integration into transmission systems is relatively easy; therefore, this diplexer type is

seen as a universal device for ECRH and (in-line) diagnostics applications.

A diplexer test model (MQ IV) is in production, which is compatible with the ITER ECRH

system (170 GHz, 63.5 mm waveguide, vacuum tight casing), with the final goal of high-

power tests at the 170 GHz gyrotron facility at JAEA in Naka, Japan. In the paper, the design

and technological issues are discussed, and first low-power measurements of the transmission

functions will be presented.

References:

[1] W. Kasparek et al., Fusion Sci. Technol. 59 (2011) 729 – 741.

[2] J. Stober et al., this conference.

[3] H. van den Brand et al., this conference.

ABSTRACT

Sensitive millimeter wave diagnostics are vulnerable to stray radiation from gyrotrons applied in Electron Cyclotron Resonance Heating (ECRH) systems at thermonuclear fusion experiments. The output power of modern gyrotrons is typically in the megawatt range with pulse lengths from several seconds to cw. This means that even a small fraction of the total injected power has the potential to destroy millimeter wave diagnostic receivers. The stray radiation can originate from non-perfect coupling to the plasma (e.g. wrong polarization of the injected millimeter wave beam) or reflections at plasma density cutoffs. There are also heating schemes with incomplete absorption of the injected millimeter wave power. For modern ECRH systems using frequency step-tunable gyrotrons, filters with more than one stop-band are required. At the same time the stop band must be rather wide (several hundreds of MHz) in order to cope for the gyrotron frequency chirp, especially at the beginning of the gyrotron pulse due to cavity heating and expansion. It must also include slightly different oscillating frequencies in systems using more than one gyrotron. Both requirements are difficult to fulfill using the available filter technology [1]. A filter based on an oversized circular corrugated waveguide with a corrugation period satisfying the Bragg condition can provide several defined stop bands with steep frequency slopes and defined width. Such a filter was successfully built and tested in Ka-band [2]. We applied the same design principle for a filter which will protect a new inline Electron Cyclotron Emission (ECE) diagnostic at ASDEX Upgrade [3] where a multi-frequency ECRH system with several two-frequency gyrotrons (105 and 140 GHz) is in operation. The filter consists of an oversized circular waveguide section with radial corrugations and a smooth-wall nonlinear taper at each end to connect the oversized waveguide section to standard circular D-Band waveguides. The corrugation was designed to couple the incident fundamental TE11 mode to the TM12 mode at 105 GHz and to the TM13 mode at 140 GHz. Both modes are trapped in the oversized waveguide section and converted back to the reflected TE11 input mode after reflection at the non-linear taper. This way, the combination of taper and Bragg reflector creates almost total reflection of the TE11 input mode at these Bragg resonances. The Bragg reflector was manufactured by SWISSto12 SA, where the required mechanical accuracy of ± 5 µm could be achieved by stacking stainless steel rings, cut by electric discharge machining, in a high precision guiding pipe [4]. The two smooth-wall tapers were fabricated by electroforming. Calculations were done using the mode matching method (scattering matrix calculations). Measurements were performed with the Rohde&Schwarz Vector Network Analyzer (ZVA24).

REFERENCES

[1] P. Woskov, “Notch Filter Options for ITER Stray Gyrotron Radiation”, Proceedings 13th International Symposium on Laser Aided Plasma Diagnostics, Takayama, Japan, 2007.

[2] D. Wagner et al., J. Infrared Milli Terahz Waves (2011) 32:1424–1433. [3] Bongers et al., Proceedings 17th Joint Workshop on Electron Cyclotron Emission and Electron Cyclotron Resonance Heating, Deurne, The Netherlands,

EPJ Web of Conferences, Volume 32, id.03006 [4] E. De Rijk et al., Rev. Sci. Instrum. 82, 066102 (2011).

D.Wagnera, W. Bongersd, W. Kasparekb, F.Leuterera, F.Monacoa, M.Münicha, H.Schütza, J.Stobera, M.Thummc, H. v.d.Brandd

a Max-Planck-Institut für Plasmaphysik, EURATOM-IPP, Boltzmansstr.2, D-85748 Garching b Institut für Grenzflächenverfahrenstechnik und Plasmatechnologie, Univ. Stuttgart, D-70569 Stuttgart, Germany

cKarlsruhe Institute of Technology, Association EURATOM-KIT, Institut für Hochleistungsimpuls- und Mikrowellentechnik, D-76021, Karlsruhe, Germany

bDutch Institute for Fundamental Energy Research, EURATOM-DIFFER, Edisonbaan 14, NL-3439 MN Nieuwegein, The Netherlands

A Multifrequency Notch Filter for Millimeter Wave Plasma Diagnostics Based on Photonic Bandgaps in Corrugated Circular Waveguides